ORAU TEAM

Dose Reconstruction

Project for NIOSH

Oak Ridge Associated Universities I Dade Moeller & Associates I MJW Corporation

Page 1 of 47

Document Title:

An Exposure Matrix for W.R. Grace and 

Company in Erwin, Tennessee

Document Number: ORAUT-TKBS-0043

Revision: 00

Effective Date: 02/14/2006

Type of Document: TBD

Supersedes: None 

Subject Experts: Paul J. Demopoulos, Donald Palmrose, and Milton Gorden 

Document Owner

Approval: Signature on File Approval Date: 01/26/2006

Paul J. Demopoulos, TBD Team Leader

Approval: Signature on File Approval Date: 01/26/2006

Judson L. Kenoyer, Task 3 Manager

Concurrence: Signature on File Concurrence Date: 02/06/2006

Kate Kimpan, Project Director

Approval: Signature on File Approval Date: 02/14/2006

James W. Neton, Associate Director for Science

 New Total Rewrite Revision Page Change

FOR DOCUMENTS MARKED AS A TOTAL REWRITE, REVISION, OR PAGE CHANGE, REPLACE THE PRIOR 

REVISION AND DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION.
Document No. ORAUT-TKBS-0043 Revision No. 00 Effective Date: 02/14/2006 Page 2 of 47

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION 

NUMBER DESCRIPTION

02/14/2006 00 New technical basis document for the W. R. Grace and Company in 

Erwin, Tennessee. First approved issue. Training required: As 

determined by the Task Manager. Initiated by Paul J. Demopoulos.
Document No. ORAUT-TKBS-0043 Revision No. 00 Effective Date: 02/14/2006 Page 3 of 47

TABLE OF CONTENTS

SECTION TITLE

Acronyms and Abbreviations ...............................................................................................................

PAGE

6

1.0 Introduction ..............................................................................................................................8

2.0 Site Description, History, and Process .....................................................................................8

2.1 Process Description......................................................................................................9

 2.1.1 Production of Uranium Metal and Uranium Metal Alloys

 Enriched up to 12% Uranium-235 .................................................................. 10

 2.1.1.1 Conversion of Uranium Hexafluoride to Uranium Metal.................... 10

 2.1.1.2 Reduction of Uranium Tetrafluoride to Uranium Metal...................... 11

 2.1.1.3 Pickling ............................................................................................ 11

 2.1.1.4 Other Metal Products....................................................................... 11

 2.1.1.5 Packaging and Shipping .................................................................. 11

2.1.2 Scrap Recovery Operations ............................................................................ 11

 2.1.3 Highly Enriched Uranium from Scrap Recovery .............................................. 12

 2.1.3.1 Receiving and Storage ...................................................................... 12

 2.1.3.2 Sampling ........................................................................................... 12

 2.1.3.3 Head-End Operations ........................................................................ 12

 2.1.3.4 Final Process Steps........................................................................... 12

 2.1.4 Production of Uranium Oxide Mixed with Thorium Oxide ................................... 

 and Zirconium Oxide ...................................................................................... 13

 2.1.4.1 Receiving and Storage...................................................................... 13

 2.1.4.2 Solution Concentration...................................................................... 13

 2.1.4.3 Precipitation...................................................................................... 13

 2.1.4.4 Drying and Grinding.......................................................................... 13

 2.1.4.5 Calcination to Uranium Dioxide......................................................... 13

 2.1.4.6 Blending ........................................................................................... 13

 2.1.4.7 Diluent Addition and Blending........................................................... 13

 2.1.4.8 Binder Addition ................................................................................. 13

 2.1.4.9 Lubricant Addition and Pressing ....................................................... 14

 2.1.4.10 Binder Removal and Sintering .......................................................... 14

 2.1.4.11 Physical Measurement and Grinding ................................................ 14

 2.1.4.12 Tube Loading.................................................................................... 14

 2.1.5 Production of Mixed Oxide Fuel ...................................................................... 14

 2.1.6 Uranium Hexafluoride Cylinder Washing......................................................... 14

3.0 Estimation of Internal Exposure.............................................................................................. 14

3.1 Uranium Exposures ....................................................................................................15

 3.1.1 Uranium Air Sampling ..................................................................................... 16

 3.1.2 Enriched and Recycled Uranium..................................................................... 20

 3.1.3 Uranium Bioassay........................................................................................... 21

3.2 Plutonium Exposures..................................................................................................22

 3.2.1 Plutonium Organ Dose Calculations................................................................ 23

 3.2.2 Plutonium Bioassay ........................................................................................ 23

3.3 Thorium Exposures ....................................................................................................26

3.4 Uranium and Plutonium Dioxide- MOX, Mixed Oxide Fuel Exposures ........................ 26

3.5 Americium ..................................................................................................................27

3.6 Occupational Internal Dose Reconstruction Summary And Assumptions - 
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 RESERVED................................................................................................................27

4.0 Estimation of External Exposure ............................................................................................ 27

4.1 Recorded Occupational Exposure and External Dosimetry......................................... 27

 4.1.1 NRC Annual Reports....................................................................................... 28

 4.1.2 Penetrating Dose Determination and Deep Dose 

 Calculations.................................................................................................... 32

 4.1.3 Non-penetrating Dose Determination and Skin Dose 

 Calculations..................................................................................................... 32

 4.1.4 Neutron Dose Assessment.............................................................................. 34

4.2 Enriched Uranium Source Term Information...............................................................34

4.3 Surface Contamination ...............................................................................................34

4.4 Plutonium Operations ................................................................................................. 40

4.5 Medical X-rays............................................................................................................41

4.6 Uranium-233...............................................................................................................41

4.7 Occupational External Dose Reconstruction Assumptions and Summary................... 41

5.0 Estimation of Exposure to Residual Activity – RESERVED .................................................... 43

6.0 Environmental Dose - reserved.............................................................................................. 43

References ........................................................................................................................................44
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LIST OF TABLES

TABLE TITLE

2-1 W.R. Grace and NFS AEC weapons related work time period and locations..........................

PAGE

10

3-1 Fuel types, chemical form, isotope and enrichment of W.R. Grace process 

material..................................................................................................................................15

3-2 Uranium source term information ........................................................................................... 16

3-3 Occupational exposures for 93% uranium-aluminum alloy recovery....................................... 16

3-4 Average breathing-zone samples for 93% uranium-aluminum alloy recovery......................... 17

3-5 Average general air-sampling concentrations for 93% uranium-aluminum 

alloy recovery.........................................................................................................................17

3-6 Occupational exposures for 3.6% uranium recovery from slag ............................................... 18

3-7 Average breathing-zone samples for 3.6% uranium recovery from slag................................. 18

3-8 Average general air-sampling concentrations for 3.6% uranium recovery from 

slag ........................................................................................................................................19

3-9 Estimated uranium inhalation intake rates based on time weighted air 

concentrations measured during uranium recovery operations............................................... 20

3-10 Estimated uranium ingestion rates based on time weighted air concentrations 

measured during uranium recovery operations ...................................................................... 20

3-11 Intakes of contaminants in recycled uranium as fractions of uranium intake........................... 21

3-12 Uranium urinalysis recorded results for claims submitted as of 10/01/05 in 

dpm/L.....................................................................................................................................22

3-13 Activity composition of Hanford reference fuel-grade plutonium mixture (12%) ...................... 23

3-14 Organ dose for 1 yr for Hanford reference 10-yr-aged fuel-grade plutonium 

mixture (12%) and Hanford reference 5-yr-aged fuel-grade plutonium mixture 

(6%) for an annual chronic intake of 6.24 × 105 pCi/yr and absorption type M........................ 24

3-16 Plutonium urinalysis results for claims.................................................................................... 25

4-1 Minimum detectable level (MDL) and maximum potential missed photon or 

beta dose...............................................................................................................................28

4-2 Dosimetry for external whole-body, wrist, and extremity exposures ....................................... 29

4-3 Annual occupational radiation exposures for 1982 to 1994 .................................................... 29

4-4 Annual occupational TEDE for 1995 to 2002.......................................................................... 31

4-5 Annual organ doses due to exposure to a 5-by-5 stacked array of birdcages......................... 34

4-6 Uranium activity fractions ....................................................................................................... 36

4-7 External exposure from thorium surface contamination.......................................................... 36

4-8 External exposure from 233U/thorium surface contamination................................................... 37

4-9 External exposure from 93%-enriched uranium surface contamination .................................. 37

4-10 External exposure from 3.5%-enriched uranium surface contamination ................................. 38

4-11 External exposure from DU surface contamination................................................................. 38

4-12 External exposure from plutonium surface contamination ...................................................... 39

4-13 External exposure from MOX surface contamination.............................................................. 39

4-14 Plutonium isotopic content and specific activity ...................................................................... 40

4-15 Annual organ doses due to the assumed annual diagnostic chest X-ray................................ 41

4-16 Operational external exposure summary................................................................................ 42

LIST OF FIGURES

FIGURE TITLE

2-1 Location of Erwin, Tennessee ..................................................................................................

PAGE

9

2-2 Plant site at W.R. Grace/NFS...................................................................................................9
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ACRONYMS AND ABBREVIATIONS

ADU ammonium diuranate

AEC U.S. Atomic Energy Commission

CDE committed dose equivalent

CFR Code of Federal Regulations

cm centimeter

d day

DCF dose conversion factor

DOE U.S. Department of Energy

dpm disintegrations per minute

DU depleted uranium

EEOICPA Energy Employees Occupational Illness Compensation Program Act of 2000

EU enriched uranium

g gram

GSD geometric standard deviation

HASL Health and Safety Laboratory

HEU highly enriched uranium

hr hour

ICRP International Commission on Radiological Protection

IMBA Integrated Modules for Bioassay Analysis

IREP Interactive RadioEpidemiology Program

keV kilo electron volt, 1,000 electron volts

kg kilogram

L liter

LEU low-enriched uranium

m meter

MDA minimum detectable activity

MDC minimum detectable concentration

MDL minimum detectable level

MeV mega electron volt, 1 million electron volts

ml milliliter

MOX mixed oxide

mR milliroentgen

mrad millirad

mrem millirem

nCi nanocurie

NFS Nuclear Fuels Services

NIOSH National Institute for Occupational Safety and Health

NRC U.S. Nuclear Regulatory Commission

pCi picocurie
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SEFOR South-West Experimental Fast Oxide Reactor

TBD technical basis document

U.S.C. United States Code

yr year

μCi microcurie

μg microgram

μm micrometer
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1.0

Technical basis documents (TBDs) and site profile documents are general working documents that 

provide guidance concerning the preparation of dose reconstructions at particular sites or categories 

of sites. They will be revised in the event additional relevant information is obtained about the affected 

site(s). These documents may be used to assist the National Institute for Occupational Safety and 

Health (NIOSH) in the completion of the individual work required for each dose reconstruction.

INTRODUCTION

In this document the word “facility” is used as a general term for an area, building or group of buildings 

that served a specific purpose at a site. It does not necessarily connote an “atomic weapons employer 

[AWE] facility” or a “U.S. Department of Energy [DOE] facility” as defined in the Energy Employees 

Occupational Illness Compensation Program Act of 2000 [EEOICPA; 42 U.S.C. §7384l(5) and (12)]. 

EEOICPA, as amended, provides for employees who worked at an AWE facility during the contract 

period and/or during the residual contamination period. 

Employment at an AWE facility is categorized as either (1) during the contract period (i.e., when the 

AWE was processing or producing material that emitted radiation and was used in the production of 

an atomic weapon), or (2) during the residual contamination period (i.e., periods that NIOSH has 

determined there is the potential for significant residual contamination outside of the period in which 

weapons-related production occurred). For contract period employment, all radiation exposures must 

be included in dose reconstructions. For residual contamination period employment, only the radiation 

exposures defined in 42 U.S.C. §7384n(c) (4) [i.e., radiation doses received from DOE/U.S. Atomic 

Energy Commission (AEC) -related work] must be included in dose reconstructions. This site profile 

covers only exposures from nuclear weapons-related work. Exposures from non-weapons related 

work, if applicable, will be covered elsewhere.

This site profile provides specific information concerning documentation of historical practices at the 

W.R. Grace and Company plant. If the dose reconstructor cannot determine whether the work 

performed was AEC weapons related work or from non-AEC related work then the dose reconstructor 

should assume that all of the work performed was from AEC weapons related work to ensure 

claimant-favorability.

2.0

The Davison Chemical Company, a division of W.R. Grace, began processing radioactive materials in 

late 1957 at the site of the current Nuclear Fuels Services (NFS) facility near Erwin, Tennessee. (NFS 

2005) In 1964, NFS was formed by the merger of W.R. Grace, and American Machine and Foundry 

Company. Ownership of NFS transitioned from Getty Oil to Texaco as part of a stock buyout of Getty 

Oil. In 1987, NFS Services, Ltd., a private Atlanta-based limited partnership purchased NFS, Inc., 

from Texaco, Inc. Figure 2-1 shows the location of Erwin, Tennessee. Figure 2-2 shows the principal 

site infrastructure for the NFS facility as of 1986 (Congress 1986).

SITE DESCRIPTION, HISTORY, AND PROCESS

Regulatory authority over operations was originally under the AEC (1954 to 1974) and transitioned to 

the U.S. Nuclear Regulatory Commission (NRC) (1975 to present) under Special Nuclear Material 

License No. SNM-124, as amended. 

The principal operation at the site has been to convert highly enriched uranium (HEU) and lowenriched uranium (LEU) from UF6 to a product that meets customer requirements. Thorium, depleted 

uranium (DU), 233U, and plutonium have also been processed at various times to oxides or metals with 

subsequent processing into the form necessary for the manufacture of nuclear fuel. Specifically, the 
Document No. ORAUT-TKBS-0043 Revision No. 00 Effective Date: 02/14/2006 Page 9 of 47

facility at one time processed ThO2 mixed with 233U to make the light-water breeder reactor fuel for the 

Shippingport Reactor. Plutonium and DU mixed oxide (MOX) fuel for the South-West Experimental 

Figure 2-1. Location of Erwin, Tennessee. 

051905_01_TB Source: Erwin Nuclear Fuels Services. Hearing Before the Subcommittee on Energy Conservation and

Power of the Committee on Energy and Commerce House of Representatives. Serial No. 99-178.

Burial Sites

Waste Retention

Ponds

Nearest Residence

Clinchfield Railroad Main Track

Area Inside Security Fence: 21.2 Acres

Total Plant Site Area: 57.8 Acres

2

3

1

DOE Property

Security Fence

Barbed Wire Fence

Figure 2-2. Plant site at W.R. Grace/NFS. 

Fast Oxide Reactor (SEFOR) was fabricated in Building 234 with laboratory capabilities in Building 

110; both operations ceased in 1970. Other operations were associated with the conversion 

operations such as scrap recovery operations (uranium and other nuclear fuel material) and cleaning 

and certification of empty cylinders used to transport low-enriched UF6. Table 2-1 lists the 

approximate processing history of the site.

2.1 PROCESS DESCRIPTION

Brief summaries for the principal operations are given below and additional details can be found in the 

series of Feasibility Reports provided in the cited references. Inherent in all of the operations is 

nuclear criticality safety that governs not only the operations and storage but also the movement of 

material within the facility. Nuclear criticality safety is maintained at the facility through the control of 

one or more of the conventional parameters of geometry, mass, concentration, and control of neutron 

interaction between subcritical units. The standard administrative policy at this and other similar 

facilities is to control two such parameters whenever possible. For birdcage units, the two control 
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Table 2-1. W.R. Grace and NFS AEC weapons related work time period and locations. 

Operations

Period of AEC Weapons 

Related Work Building Location

Thorium (metal & oxide) Late 1957 to 1970 110C, 110D, 111, 130, 234B, 

234C, 310

Uranium metal (HEU & DU) Late 1957 to 1970 110, 110E, 111, 130, 135, 234, 

301

Low-enriched UO2 Early 1960 to 1970 301, 110E, 111, 130, 135, 301

233U fuel (sometimes mixed with 

ThO2)

1961 -1970 234B, 234C, 110C, 110D

Plutonium fuel and MOX 1966 to 1970 234A, 234B, 234C, 110C, 110D

HEU scrap recovery Late 1957 to 1968

1968 through 1970a 

130

220, 230, 233

HEU Fuel Early 1966 to 1970a 100, 105, 120, 131, 132/133, 220, 

233, 300, 302, 303, 304, 310, 330

(Congress 1986, p. 111) 301 (from 

1982) (Congress 1986, p. 132)

LEU Scrap Recovery 1960 to 1970a (AEC 1961), or 

(Congress 1986)

111, 220, 230, 233

LEU cylinder washing 1970a 111,130

Sources: AEC (1961), Congress (1986, pp. 111-114 and 127-143).

a. The end of 1970 is when AEC-weapons related operations ended.

parameters are geometry (birdcage dimensions) and a 235U mass limit that can be placed within the 

birdcage. Examples of bird cages that were used included: U-233 storage bird cage that used a 

shielded 55-gallon drum that contained inserts with spacers; UF6 cylinders storage bird cage that used 

a 2” × 2” frame made of metal that held the 5” UF6 cylinder in place at its center with a total dimension 

of 52” high and 30 “ wide; and a U3O8 storage bird cage that used a 2” × 2” metal frame made of metal 

that held a 10 3⁄4” × 12” cylinder in the center with a total dimension of 36” high and 30 5/8 “. A review 

of the available literature shows that no criticality accidents have occurred during W.R. Grace and 

later NFS operations.

2.1.1

The UF6 was supplied in approved standard cylinders and received in approved packaging such as 

birdcages. The cylinders were check-weighed and placed into storage in special concrete cells or 

birdcages in one of the warehouse buildings.

Production of Uranium Metal and Uranium Metal Alloys Enriched up to 

12% Uranium-235

2.1.1.1 Conversion of Uranium Hexafluoride to Uranium Metal

The overall process involves vaporization, reduction to UF4, reduction to uranium metal, pickling, 

processing into other metal products, packaging, and shipment. The following information is from 

Feasibility Report for the Production of Uranium Metal and Uranium Metal Alloys Enriched Up To 12% 

235U (Housholder 1963a).

UF6 was received in solid form in cylinders and had to be vaporized to transfer it to the UF6 to UF4

reactor (the 6 to 4 unit). Vaporization was accomplished by heating the UF6 cylinders in an electric 

oven. Up to six cylinders could be placed in to the oven for processing.

The UF6 was piped into the 6 to 4 unit where it was reduced with hydrogen. The solid UF4 powder 

dropped into a product hopper where it was metered into safe diameter product cans. The off gas 

flowed through two cyclone separators where any entrained UF4 dust was collected in additional 
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product cans. The off gas was then filtered to remove the last traces of UF4 dust and was scrubbed 

with KOH to remove HF vapors. The gas, free of uranium and acid, was vented to the atmosphere 

through a flame arrester where excess hydrogen was burned off.

2.1.1.2 Reduction of Uranium Tetrafluoride to Uranium Metal

The UF4 was weighed into a reduction batch and was blended with a reducing agent such as 

magnesium metal. The charge was then heated under vacuum in an induction furnace to form a 

uranium metal derby. After cooling, the uranium metal derby was broken out and separated from the 

slag. The slag was packaged for scrap recovery where any uranium remaining in the slag was 

recovered.

2.1.1.3 Pickling

The uranium derby was pickled in acid to remove adherent slag and scale. The pickle solution was 

sent to scrap recovery for recovery of uranium. The pickled derby was then sampled for impurities.

2.1.1.4 Other Metal Products

As necessary, the derbies were broken into smaller pieces before shipment on a large hydraulic 

press. On other occasions, the derbies were remelted and cast into various shapes such as slugs, 

rods, and plates. The uranium could also be alloyed with other metals during remelting and casting.

Melting was done in a large vacuum induction furnace. Cast pieces could be pickled as previously 

described.

2.1.1.5 Packaging and Shipping

All enriched uranium (EU) metal products were packaged for shipment in approved birdcages. The 

products could be stored on site for a time before shipping to the customer. 

2.1.2

This operation was designed to recover the scrap uranium from the manufacturing processes. The 

process steps were assumed to be carried out in a similar manner for scrap with either HEU or LEU 

(Katine 1960). The exception between handling the different enrichment levels would be in the 

dissolution and filtration steps. There was potential for different end processes that were adjusted 

based on the final material form desired by either the customer (for scrap recovery from outside 

sources) or how the facility was going to reinsert the recovered uranium in the fuel fabrication 

process. 

Scrap Recovery Operations

Dependent of whether the material was greater or less than 2% enrichment, two different batch sizes 

and dissolver tanks were used. For scrap material equal to less than 2% enrichment, a maximum 

batch of 87.95 kg uranium (220 pounds) of UO2 is dissolved in a 500-gallon capacity dissolver tank.

For all other enrichments, a 20-gallon dissolver tank was employed and the batch would contain a 

maximum of 400 g of 235U. 

Of note for this process, the memorandum attached to a Feasibility Report for HEU scrap recovery 

specifically states “the chopping and sawing of solid metal plates, billets, and rods, and the operations 

of grinding, screening, and blending of the resultant oxides from the calcinations step, could be 

possible dust sources. Air samples should be obtained at these operations and evaluated at the start 
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of operations in order to ascertain the effectiveness of the existing ventilation (Glauberman 1962).” A 

summary of the HEU scrap recovery follows (Housholder and Runion 1962).

2.1.3

2.1.3.1 Receiving and Storage

Highly Enriched Uranium from Scrap Recovery

All scrap would be received in approved shipping containers and birdcages and sent to storage 

warehouses to await processing.

2.1.3.2 Sampling

Before processing, each container was inspected for the purpose of deciding what head-end steps 

could be necessary before the scrap could be dissolved. Also at this time, samples were taken for 

uranium assay for the purpose of determining batch sizes.

2.1.3.3 Head-End Operations

Due to the many types of scrap received, it was often necessary to pretreat the scrap before 

dissolution and subsequent extraction. Metal chips and turnings, for instance, were routinely received 

stored under oil, which must be removed before dissolution. This was done by draining the oil off the 

metal in wire baskets. The metal was then washed with solvents and dried.

Solid metal plates and rods were chopped or sawed into smaller pieces to facilitate handling and 

weighing into batches. Combustible wastes such as filters, sponges, grinder sludge, sweepings, etc., 

were calcined to reduce their bulk and to remove hydrogenous and carbonaceous materials, such as 

water and oil. After calcining, the resultant oxides were ground and screened; material not passing the 

screen was recycled to the grinder. The oxides were then blended to ensure homogeneity for 

sampling for uranium assay from which accountability and batch sizes could be determined.

Oxide pellets were ground and screened and could be calcined and blended to facilitate dissolution.

2.1.3.4 Final Process Steps

A review of the Feasibility Reports showed there were different final process steps that appear to 

depend on the final material form needed either to be reinserted into the facility process or to be 

shipped back to the customer. The extracted uranium liquid (also known as O.K. Liquor) could be 

converted into a solid through either a boil-down and crystallization process (Housholder and Runion 

1962) or by precipitation and filtration followed by calcination and blending (AEC 1959). A summary of 

both final process steps follows:

• Boil-down and Crystallization

•

: The pure uranium solution could be boiled down and 

evaporated to produce uranyl salts, which were packaged for shipment. At times, the 

concentrated solution from the boil-down step was packaged in bottles for shipment as a 

solution.

Precipitation/Filtration and Calcination/Blending: The extracted uranium liquid was precipitated 

by batch process with ammonium hydroxide in an approved container with an agitator. The 

precipitated slurry was transferred to the filtration hood for filtration with a large Buchner 

funnel. The filter cake was loaded into metal trays and transferred to a hooded muffle furnace 

where it was calcined to U3O8. Once cooled, the material was transferred to a blending station 
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for introduction into a ball mill for grinding and blending. After ball milling, the U3O8 was 

sampled, packaged, weighed, and returned to the storage area ready for shipment. 

2.1.4

This line of production (233UO2/ThO2 and 233UO2/ZrO2) was in operation from approximately 1961 to 

1969 for the light-water breeder reactor fuel of the Shippingport Reactor. This process was more 

involved than other process lines with eleven separate steps to produce the 233U fuel (Housholder 

1963b). A brief summary of the process follows.

Production of Uranium Oxide Mixed with Thorium Oxide and Zirconium Oxide 

2.1.4.1 Receiving and Storage

233U was received as a uranyl nitrate solution in an approved shipping cask. A typical receipt shipment 

was approximately 7 kg of 233U stored in storage columns waiting processing.

2.1.4.2 Solution Concentration

The received uranyl nitrate solution had a concentration of approximately 150 g/L 233U. Before 

precipitation, a higher concentration was required and was obtained through evaporation.

2.1.4.3 Precipitation

The uranyl nitrate solution was measured into safe batches and the uranium was precipitated by the 

addition of a precipitating agent such as NH4OH.

2.1.4.4 Drying and Grinding

The resultant precipitate was dried at a low temperature and ground to a fine powder.

2.1.4.5 Calcination to Uranium Dioxide

The dried ground uranium precipitate was calcined to UO2 in a continuous muffle furnace under a 

hydrogen atmosphere. The resultant oxide was stored in birdcages to await blending.

2.1.4.6 Blending

To ensure homogeneity of the oxide, the precipitation batches were accumulated and blended before 

addition of the diluents (ThO2 or ZrO2). After a homogenous UO2 blend was obtained, it was sorted in 

birdcages to await blending with the diluents.

2.1.4.7 Diluent Addition and Blending

Safe batches of UO2 and diluents were weighed out and blended together in a twin-shell blender. To 

ensure thorough mixing, each batch was also ball milled.

2.1.4.8 Binder Addition

A binder was mixed with the oxide blend, and the wet mixture was granulated and dried. After drying, 

the granules were broken up by screening.
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2.1.4.9 Lubricant Addition and Pressing

To improve pressability, a die lubricant was added to the granules and blended in. The oxides were 

then compacted into small pellets on a 40-ton press.

2.1.4.10 Binder Removal and Sintering

The resultant pellets were loaded into trays and heated in an oven to drive off the binder. They were 

then sintered in a continuous muffle furnace under a hydrogen atmosphere.

2.1.4.11 Physical Measurement and Grinding

The sintered pellets were inspected and measured to see that they met customer’s requirements.

Before packaging, the sintered pellets could have required grinding to the correct diameter If grinding 

were necessary, the pellets were cleaned by washing them in water, drying them, and re-measuring 

them.

2.1.4.12 Tube Loading

The finished pellets were stacked, weighed, and loaded into zircalloy tubes. The tubes were welded 

closed and loaded into shielded 55-gallon drum birdcages for temporary storage and eventual 

shipment to the customer. 

2.1.5

Documentation of the process steps for the production of MOX fuel (combination of PuO2) with 

low-enriched UO2) for the SAFOR reactor at the facility was not available. However, other DOE sites 

have performed similar operations. From these documents, production of MOX fuel would have 

followed similar process steps for the production of 233UO2/ThO2 and 233UO2/ZrO2 fuel from the 

blending step on to tube loading. The process steps were completely within shielded and filtered 

gloveboxes for worker health and safety. It is assumed that the plutonium sent to the facility was in 

approved shipping packages as PuO2 powder in sealed cans. The final product would have had 

approximately 5% 239Pu in fuel pin assemblies loaded into shielded, 55-gallon drum birdcages for 

temporary storage and shipment to the customer. 

Production of Mixed Oxide Fuel

2.1.6

UF6 cylinder washing was performed in the Building 200 complex to recover uranium in a ventilated 

glovebox using water or steam. The removed wash solution was transferred to the HEU scrap 

recovery process to recover the uranium.

Uranium Hexafluoride Cylinder Washing

3.0

The primary sources of internal radiation exposure at W.R. Grace were uranium, plutonium, or 

thorium dust produced from the manipulation and chemical processing of those materials during 

recovery and fuel fabrication processes. Uranium enrichment levels included depleted, natural, low 

(3.5%), and high (93%). There was one indication of recycled uranium (RU) processing in the case of 

a recycled 233U pellet (Housholder 1963b). Table 3-1 lists the various enrichments and chemical forms 

of the processed radionuclides. 

ESTIMATION OF INTERNAL EXPOSURE
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Table 3-1. Fuel types, chemical form, isotope and enrichment of W.R. Grace process material. 

Radionuclide or fuel

Chemical form and 

solubility type(s) Isotope Enrichment

Uranium UF6, UO2F2, & UO2(NO3)2

(F)

UO3 & UF4 (M)

U3O8 & UO2, (S)

233U 234U 235U 236U 238U

DU, Natural, 

LEU (3.5%),

HEU (93%)

Thorium ThO2 (S) 230Th (234U) 231Th (235U) 232Th (236U) 234Th (238U) 229Th (233U) Natural 

Plutonium PuO2 (S) 238Pu 0.64%, 239Pu 2.06%, 

240Pu 1.07%, 241Pu 95.4%, 

241Am 0.86% (% Activity) 

Fuel Grade

Aged 10 yr

Technetium or other 

transuranic elements

Same as the Th, U or Pu 

matrix

99Tc, 237Np NA 

MOX PuO2 /UO2 (S) 20% PuO2 and 80% UO2

by weight

About 3.5% 235U 

233U fuel ThO2 /233UO2 or ZrO2

/

233UO2 (S)

20% UO2 and 80% ThO2 or 

ZrO2 by weight

HEU (Likely > 20%)

3.1 URANIUM EXPOSURES

For a given uranium process, the mass of (long-lived) uranium released to air does not change 

because of enrichment (ORAUT 2005a). 

Uranium air sampling is further discussed in Section 3.1.1. 

In general, it is not known how much of any specific enrichment versus another to which a worker 

could have been exposed. The uranium source term information in Table 3-2 can be used only if the 

information on the uranium enrichments that the worker processed is known. It is most claimantfavorable to assume the highest known processed enrichment.

Human and animal studies have indicated that oxides of uranium can be very insoluble (ICRP 1995), 

which indicates absorption type S. Other in vitro dissolution studies of compounds found at uranium 

facilities have shown that the oxides of uranium have exhibited moderate solubility (ORAUT 2005a), 

which suggests absorption type M. Because there was no specific information on the solubility of 

aerosols produced at W.R. Grace, this analysis assumed that both types M and S were available. The 

selection of absorption type should depend on the organ of interest. For example, absorption type S 

should be selected for respiratory tract dose calculations. Dose reconstructions should assume 

International Commission on Radiological Protection (ICRP) Publication 66 default parameters for 

particle deposition (ICRP 1994a).

ICRP (1994) lists UF6, UO2F2, and UO2(NO3)2 (uranyl nitrate) as type F; UF4 and UO3 as type M; and 

U3O8 and UO2 as type S. The chemical form and the enrichment varied over time at W.R. Grace. The 

manufacture of uranium occurred in most of the buildings in W.R. Grace with the exception of 

Buildings 234A, 234B, and 234C, where plutonium was processed.
Document No. ORAUT-TKBS-0043 Revision No. 00 Effective Date: 02/14/2006 Page 16 of 47

Table 3-2. Uranium source term information. 

Uranium source term Reference

Specific activity

pCi/μg

Activity fractions 234U 235U 236U 238U

Natural uranium IMBAa 0.683 0.489 0.023 - 0.489

93.% IMBAa 68.1 0.968 0.030 0.002 0.0003

3.5% IMBAa 2.20 0.818 0.034 - 0.147

2% HPSb 1.616 0.648 0.041 0.0009 0.311

Typical DU IMBAa 0.402 0.155 0.011 0.0005 0.834

RU (1% 235U) Hanfordc 0.910 0.563 0.023 0.048 0.365

Uranium source term Reference

Specific activity

pCi/μg

Specific constituent activity in mixture 

(μCi/g, nCi/mg, or pCi/μg)

Natural Uranium IMBAa 0.683 0.334 0.016 - 0.334

93.% IMBAa 68.1 65.9 2.04 0.136 0.020

3.5% IMBAa 2.20 1.80 0.075 - 0.323

2% HPSb 1.616 0.778 0.049 0.001 0.373

Typical DU IMBAa 0.402 0.062 0.004 0.0002 0.335

RU (1% 235U) Hanfordc 0.910 0.563 0.023 0.048 0.365

a. IMBA = Integrated Modules for Bioassay Analysis software

b. HPS N13.22 (ANSI 1995). 

c. ORAUT (2004a).

3.1.1

This discussion focuses on documented air-sampling data from two separate Health and Safety 

Laboratory (HASL) reports by the AEC. The first air sampling was performed at W.R. Grace during the 

recovery of 93% HEU from uranium-aluminum alloy scrap in the storage, the ceramics, and chemical 

buildings (AEC 1959). The second air sampling was performed during the recovery of 3.6% LEU from 

slag scrap (AEC 1961). Table 3-3 to 3-8 lists the results of the sampling analyses. 

Uranium Air Sampling

Table 3-3. Occupational exposures for 93% uranium-aluminum alloy recovery.

a 

Operator

Number of 

persons

Average daily weighted 

exposures in dpm/m3

Column operator 6 31

Ceramic building operator 3 45

Charge makeup 3 14

Accountability 3 15

Chemist and Technicians 8 3.1

Spectrographic Operator and Technician 3 2.5

Machine shop 5 12

Laundry 1 7

Geometric meanb 10.84

Geometric standard deviationb 2.76

95% confidence level valueb 57.63

90% confidence level valueb 39.84

50% confidence level valueb 10.84

a. Data from AEC (1959).

b. Statistics calculated using a log normal distribution and the program Crystal Ball.

The air samples consisted of collection on filters of radioactive particulates from breathing zones and 

general areas during processing. The alpha activity measured on the filter was used to determine the 

airborne alpha activity concentrations. When multiple samples at a location were collected, the AEC 

used the mean air concentration in subsequent calculations. The AEC matched air concentration 
Document No. ORAUT-TKBS-0043 Revision No. 00 Effective Date: 02/14/2006 Page 17 of 47

Table 3-4. Average breathing-zone samples for 93% uranium-aluminum alloy recovery.

a 

Operation

Number of 

samples

Average concentration 

dpm/m3

Shearing U-Al alloy for charge makeup and weighing 3 43

Charging digester with batch of U-Al alloy 2 170

Running material from digester thru filter press and column 3 50

Cleaning residue from filter press 2 19

Securing O.K. Liquor from columns 1 65

Precipitation of O.K. Liquor 2 16

Filtering ppt. on Buchner funnel 2 15

Removing filter cake-placing in tray 1 1

Transferring tray from furnace to cooling area 2 65

Transferring tray from furnace to dry box, weighing and unloading 

tray in dry box, cleaning residue tray 3 280

Digestion of organic ashes in hood 1 1

Filtration of digested organic ashes 1 65

Geometric meanb 26.65

Geometric standard deviationb 5.84

95% confidence level valueb 485.13

90% confidence level valueb 255.57

50% confidence level valueb 26.65

a. Data from (AEC 1959). 

b. Statistics calculated using a log normal distribution and the program Crystal Ball.

Table 3-5. Average general air-sampling concentrations for 93% 

uranium-aluminum alloy recovery.

a 

Area

Number of 

samples

Average concentration 

dpm/m3

Shearing and Weighing 2 9

Solvent extraction area 6 19

Chemical building - oven area 3 23

Ceramics building 9 2

Machine shop 4 16

Wet chemistry lab 3 3

Spectrographic lab 3 2

Laundry 3 6

Lunch room 3 7

Clean locker room 3 2

Contaminated locker room 3 4

Geometric meanb 5.81

Geometric standard deviationb 2.52

95% confidence level valueb 26.55

90% confidence level valueb 18.98

50% confidence level valueb 5.81

a. Data from AEC (1959).

b. Statistics calculated using a log normal distribution and the program Crystal Ball.

determinations with information about worker categories, locations, tasks, and time at each location or 

task. 

Although it is unlikely that workers would have been exposed to the same air concentrations from the 

many other processes that were conducted at NFS, this is the only air sample information currently 

available. In addition, changes in the processes and the safety controls could have resulted in both 

increased and decreased exposure for any given period. In this document an estimate of the intake 
Document No. ORAUT-TKBS-0043 Revision No. 00 Effective Date: 02/14/2006 Page 18 of 47

Table 3-6. Occupational exposures for 3.6% uranium recovery from slag.

a 

Operator

Number of 

persons

Average daily weighted 

exposures in dpm/m3

Williams Roll Mill 3 170

Digestion 3 45

Solvent extraction 3 10

ADU precipitation 3 45

ADU oxide (ceramics) 3 17

Micropulverizer-drifter 3 71

Accountability-shipping and receiving 2 22

Laundry 1 9.4

Janitor 1 37

Health physics technician 2 20

Geometric meanb 30.72

Geometric standard deviationb 2.46

95% confidence level valueb 132.11

90% confidence level valueb 95.55

50% confidence level valueb 30.22

a. Data from AEC (1961).

b. Statistics calculated using a log normal distribution and the program Crystal Ball.

Table 3-7. Average breathing-zone samples for 3.6% uranium recovery from slag.

a 

Operation

Number of 

samples

Average concentration 

dpm/m3

Operating jaw crusher 4 230

Loading top of Williams roll mill with crushed slag 3 500

Removing drum of 325 mesh MgF2 from hopper 2 70

Loading and unloading pot (oxidation furnace) with slag metal 

heavies 2 140

Charging digester with slag 3 150

Cleaning ADU from plate and frame filter press 3 130

Placing tray of ADU inside of oven 2 84

Removing tray of oxide from furnace-placing inside of glove 

box and transferring U3O8 4 97

Micropulverizing U3O8 (bag not working properly) 3 590

Geometric meanb 167.83

Geometric standard deviationb 2.12

95% confidence level valueb 578.38

90% confidence level valueb 440.08

50% confidence level valueb 167.83

a. Data from AEC )1961).

b. Statistics calculated using a log normal distribution and the program Crystal Ball.

was made by calculating the 95th percentile of the maximum average breathing-zone concentration 

for the 1961 AEC air sample results (from Table 3-7 later in this section) by assuming a lognormal 

distribution of the data in the AEC report (AEC 1961). Using Crystal ball, a lognormal distribution 

calculation yielded a value of 578 dpm/m3 with a geometric mean of 167 dpm/m3 and a geometric 

standard deviation (GSD) of 2.12. The calculated 95th-percentile air concentration was used to 

calculate upper estimates of internal exposures. The most conservative air concentrations were 

generated from the radiological task-oriented information in Table 3-7. If actual operator information 

from Table 3-3 and 3-6 were to be used, the calculated intakes would be about a factor of 4 to 10 

less. Task-oriented weighted average samples involve the highest air concentrations encountered 

because of the closest proximity to the actual work and no accounting of the time away from the 
Document No. ORAUT-TKBS-0043 Revision No. 00 Effective Date: 02/14/2006 Page 19 of 47

Table 3-8. Average general air-sampling concentrations for 3.6% uranium 

recovery from slag.

a 

Area

Number of 

samples

Average concentration 

dpm/m3

Crushing room 5 110

Digester area 6 11

Leach area 3 20

Solvent extraction area - all levels 8 6.4

Boil-down feed tank area 3 17

O. K. liquor storage 3 9

Raffinate storage area 3 9

Neutralizer tank storage 4 10

ADU precipitation tanks 3 16

ADU filter area 4 17

U3O8 transfer hood area 3 11

Micropulverizer area 4 62

Accountability room 5 21

Chemistry lab 4 4.3

Spectrographic lab 2 9.3

Research and development lab 4 6.6

Laundry 4 2.3

Locker room - clean side 2 17

Locker room - dirty side 2 45

Lunch room 4 21

Health physics room 2 1.5

Geometric meanb 12.57

Geometric standard deviationb 2.7

95% confidence level valueb 64.52

90% confidence level valueb 44.95

50% confidence level valueb 12.57

a. Data from AEC (1961).

b. Statistics calculated using a log normal distribution and the program Crystal Ball.

actual or lower exposure rate activities. The use of the task-oriented time-weighted average air 

concentrations is claimant favorable. 

As seen in Table 3-3, the uranium exposure by occupation differed with the column and ceramic 

building operators with the largest daily weighted exposures. This exposure variability between 

occupations is also seen in Table 3-6, where the Williams roll mill operators were clearly the highest 

exposed group, with the micropulverizer-drifter, digestion and ammonium diuranate (ADU) 

precipitation operators being the next highest groups.

The general air sample results in Tables 3-5 and 3-8 result in the lowest calculated intakes. This is to 

be expected because general air samples are typically lower than breathing-zone samples. Bioassay 

results are expected to be at levels between breathing-zone and general air sample results. 

The breathing rate is based on the default for light work, 1.2 m3

/hr, as indicated in ICRP Publication 

66 (ICRP 1994a). This category assumes an activity distribution of one-third sitting and two-thirds light 

exercise. The intakes in picocuries are calculated by dividing the 95th percentile of the air 

concentration (578 dpm/m3

) by 2.22 dpm/pCi and multiplying this result by the breathing rate and the 

assumed number of hours exposed at a given concentration. The organ doses are assumed to be a 

constant distribution. Several assumptions included in the intake and dose reconstruction are likely to 

be overestimating assumptions. This includes the use of a lognormal distribution, the 95% confidence 
Document No. ORAUT-TKBS-0043 Revision No. 00 Effective Date: 02/14/2006 Page 20 of 47

level concentration, the task-related as opposed to the occupation time-weighted average, and the 

assumption of constant work activity and worker exposure during the entire work period. 

Air sampling for 233U has also been identified. Airborne concentrations for enriched UO2 decladding 

and dissolution of 233U3O8 pellets from the immediate work area resulted in an average concentration 

of less than 1.8% of the maximum air concentration (1 × 10-10 μCi/ml) (Householder 1963b). This 

would result in calculations of intake less than that calculated above.

In the case where inhalation intakes are calculated from air concentrations, ingestion intakes are also 

to be considered. NIOSH (2004) indicates that the ingestion rate, in terms of pCi for an 8-hour 

workday, can be estimated by multiplying the air concentrations in pCi per cubic meter by a factor of 

0.2. The uranium ingestion rate based on air concentration of 260 pCi/m3 would be 52 pCi/workday. 

The daily inhalation and ingestion intake rate from LEU recovery is estimated from the 95 % 

confidence level air concentration listed in Table 3-7. Table 3-9 lists the inhalation intake rate per year 

and AEC operational time period. Table 3-10 lists the ingestion intake rate by year and AEC 

operational time period. Note, bioassay data should be used first and not the intakes calculated based 

on air sampling results unless the bioassay data is deficient. Most potentially exposed workers at WR 

Grace/NFS were monitored. Some reports in the 1970s exist that indicate skin contamination was a 

recurrent issue at W.R. Grace/NFS. The contamination levels indicated in the claimant records should 

be bounded by the assumption of 10 % of the skin contamination equilibrium activity levels being 

ingested as per NIOSH 2004, “Estimation of Ingestion Intakes”. A study was done at the Oak Ridge 

Gaseous Diffusion Plant to determine the intake from hand contamination. The study indicated that 

the amount of uranium transferred from the hand to the cigarette while smoking was approximately 

1% of the material on the surface of the hand. (Bailey 1958) The ingestion calculation included both 

contamination of food or drink from contaminated air settling and the transfer from contaminated 

surfaces to the hands to food or drink on a chronic basis. 

Table 3-9. Estimated uranium inhalation intake rates based on time weighted air concentrations 

measured during uranium recovery operations. 

Work period

Number 

of years

Number of potential 

AEC work hours per 

work period

Air 

concentration 

(pCi/m3)

Breathing 

rate (m3/hr) Intake (pCi)

1/1957-12/31/1970 14 28,000 260 1.2 8.73x106

1 yr 1 2,000 260 1.2 6.24x105

Table 3-10. Estimated uranium ingestion rates based on time weighted air concentrations 

measured during uranium recovery operations. 

Work period

Number 

of years

Number of potential 

AEC work days per 

work period

Ingestion rate 

pCi/workday Intake (pCi)

1/1957-12/31/1970 14 3,500 52 1.82 × 105

1 yr 1 250 52 1.30 × 104

a. Ingestion values were calculated according to NIOSH 2004.Choose the same f1-value as used for inhalation in 

accordance with NIOSH (2004).

3.1.2

Because the AEC air samples were counted with alpha detectors, which detect radioactivity rather 

than mass, there is no need to adjust measured air concentration results for assumed uranium 

enrichment. AEC air sample results taken during the recovery of HEU were below the concentrations 

measured during the recovery of LEU. Because various levels of enrichment occurred during the 

Enriched and Recycled Uranium
Document No. ORAUT-TKBS-0043 Revision No. 00 Effective Date: 02/14/2006 Page 21 of 47

processing and recovery operations of uranium, this document assumes that intakes are U234 for the 

purpose of internal dose calculation.

Only the uranium intake should be assigned for facilities with exposure only to uranium (or natural 

thorium), except when exposure was to recycled uranium (RU). For RU, the dose reconstructor 

should add the intakes from Table 3-11 (ORAUT 2005b).

Table 3-11. Intakes of contaminants in recycled 

uranium as fractions of uranium intake.

a

 

Radionuclide

Activity fraction of contaminant 

(e.g., pCi X/pCi U)

234U 1.0

238Pu 0.06

237Np 0.005

99Tc 0.4

232Th 0.02

228Th 0.02

106Ru 0.04

a. Source: Based on ORAUT 2005b. 

3.1.3

The highest recorded uranium in urine result for W.R. Grace/NFS from claimant records was 6229 

dpm/L recorded in 1967. Urinalysis for uranium started in October of 1964. Samples were sent to 

Eberline until about 1974. From about that time, on-site analysis was performed. Details of the 

analysis are not known. The recording level was 1 dpm/L at the 2-sigma or 95% confidence level as 

indicated from the review of employee documents as indicated from the NFS Health Physics 

department. (Tester 2005) This is likely the assumed decision level. The minimum detectable level 

should be twice the critical level or 2 dpm/L. Some random samples were sent off the site for fecal 

analysis of uranium and urine analysis for enriched uranium. Some bioassay records received from 

W.R. Grace/NFS have some of the radiometric uranium bioassay results with a minimum recording 

level of <10 dpm/L. Other records have bioassay results reported on two forms, one hand written 

form indicating laboratory results with actual results recorded down to 1.0 dpm/L, and another typed 

report in which the lowest values are recorded as <10 dpm/L, i.e., all values less than 10 dpm/L were 

recorded as <10 dpm/L on the typed form. The 10 dpm/L recording limit should be considered the 

MDA for results for which the actual lab recorded result is not available. When bioassay data is 

available, dose reconstructors should use bioassay data for estimating worker intakes. Intakes of the 

contaminants listed in Table 3-11 should be added based on the given ratios.

Uranium Bioassay

A summary table of the highest recorded values compiled from a review of claimant records is 

presented below in Table 3-12. Most of the highest recorded doses occurred during the AEC contract 

years of 1957 through 1970. Most of the urine results that greatly exceeded 100 dpm/L decreased to 

less than 100 dpm/L within several days. This included the highest values listed in Table 3-12. The 

highest values recorded from 1964 though 1970 were used to estimate the intake of uranium. 

Lung counts were performed from 1970 to the present for uranium as well as plutonium and 

neptunium. Lung counts are in general not as reliable as urinalysis for routine monitoring. However, 

this monitoring was routine and was used to assess routine exposures to transuranic and fission 

products and to further analyze results from accidental intakes. The MDA for 235U was about 120 μg 

as indicated from the cursory review of claimant documents, which is a reasonable default MDA 

value. Actual MDAs from workers records should be used when available.
Document No. ORAUT-TKBS-0043 Revision No. 00 Effective Date: 02/14/2006 Page 22 of 47

3.2 PLUTONIUM EXPOSURES

Many forms of plutonium were possible over the years including metal and oxide. Because not 

enough information for the recovery or manufacture of plutonium was found, the exact chemical forms 

are not known. It is possible that MOX fuel of a plutonium and thorium mixture was processed at 

some point because UO2 and ThO2 mixtures were produced on the site and plutonium production 

capacity existed at the same time. Most of the manufacture of plutonium occurred in Buildings 234A, 

234B, and 234C.

Table 3-12. Uranium urinalysis recorded results for claims submitted as of 

10/01/05 in dpm/L. 

Sample yeara 

Highest 

result

Second highest 

result Comment

1964 334 238

1965 725 467 Rate < 100 in 2 to 8 weeks

1966 1,480 122 Rate < 100 in < 1 week

1967 80 47

1968 80 76

1969 135 134 Rate < 100 in < 1 week

1970 234 112 Rate < 100 in < 1 week

1971 56 53

1972 133 63 Rate < 100 in < 1 week

1973 435 46 Rate < 100 in < 1 week

1974 66 30

1975 1,420 60 Rate < 100 in < 1 week

1976 75 35

1977 332 89 Rate < 100 in < 1 week

1978 55 21

1979 99 77

1980 2,427 2,129 Rate < 100 in < 1 week

1981 70 67

1982 6,229 5,423 Rate < 100 in < 1 week

1983 221 150 Rate < 100 in < 1 week

1984 94 68

1985 105 52 Rate < 100 in < 1 week

1986 229 127 Rate < 100 in < 1 week

1987 60 58

1988 15 10

1989 89 28

1990 19 20

1991 17 10

1992 67 63

a. All data from claims filed as of 11/30/05.

In general, plutonium oxides, carbides, and hydroxides are absorption type S; nitrates and other 

compounds are type M (ICRP 1995). Older materials, even when starting out as soluble, can have a 

tendency to oxidize when left in contact with air. Oxides, metals, and old contamination should be 

treated as type S. If nothing is known about the chemical form of plutonium, then either type M or S 

can be used to maximize the dose to the organ of concern. Americium-241 is a component of 

plutonium contamination and should be modeled in the lung the same as the plutonium matrix in 

which it has grown. In other words, the americium should be treated as absorption type S if the 

plutonium is type S (ORAUT 2004a). 
Document No. ORAUT-TKBS-0043 Revision No. 00 Effective Date: 02/14/2006 Page 23 of 47

There are essentially three types of plutonium-based material. Reactor grade, fuel grade and 

weapons grade with fuel grade falling in between reactor and weapons grade. For this section, lacking 

any specific information on the actual composition of the processed plutonium, an assumption of 

10-yr-old fuel-grade plutonium is claimant-favorable and reasonable (Table 3-16). This is noted in the 

Hanford site profile; Hanford processed much of the DOE complex plutonium . 

3.2.1

The isotopic breakdown of plutonium mixtures in this TBD assumes the composition as listed in Table 

3-13 which is from the Hanford reference fuel grade plutonium aged for 10 years. The assumption of 

10 year aged fuel should be claimant favorable. If the intake level for plutonium was set at the 

uranium value of 6.24 × 105 pCi/yr, as estimated from uranium time weighted air samples, the annual 

dose from the above 10-yr fuel-grade plutonium of a 12% mixture absorption type M would result in 

the organ doses listed in Table 3-14. Five-yr fuel-grade plutonium is included as a comparison. Table 

3-14 is for illustrative purposes only and indicates the possible maximally exposed organs. 

Plutonium Organ Dose Calculations

Table 3-13. Activity composition of Hanford reference fuel-grade 

plutonium mixture (12%).

a

Mixture 

designation

Specific 

activitya

Fraction 

of activity

Type Sb

pCi/yr

Type Mc

pCi/yr

Total activity/yr 1.50E+06 6.24E+05

238

Pu 1.58E-02 7.73E-03 1.16E+04 4.82E+03

239Pu 5.26E-02 2.57E-02 3.86E+04 1.60E+04

240Pu 2.72E-02 1.33E-02 2.00E+04 8.30E+03

241Pu 1.91E+00 9.34E-01 1.40E+06 5.83E+05

242Pu 3.93E-06 1.92E-06 2.88E+00 1.20E+00

241Am 3.89E-02 1.90E-02 2.85E+04 1.19E+04

a. Source: ORAUT (2004a).

b. Calculated form urinalysis data.

c. Calculated form Uranium air sample data.

3.2.2

Plutonium-239 was analyzed in urine from about 1967 to 1980. Some Pu bioassay results are also 

available from the 1980s and 1990s. Some bioassay records indicate that the later results are from 

decommissioning work. The minimum detectable concentration (MDC), analytical methodology, or 

frequency is not known. The lowest positive reported level was 0.03 dpm/L (0.014 pCi/L) on February 

5, 1967, by NFS. The highest reported level was 2.04 dpm/L (0.92 pCi/L) on October 03, 1967, by 

NFS. Forty six of the 81 recorded values were zero. Thirty five of the 81 recorded values are listed in 

Table 3-15. From the worker internal dose records, the frequency of the bioassay could have been 

about twice per year but no pattern could be determined. The years of plutonium exposures likely 

occurred from 1967 though February of 1973. (Claimant records and Congress 1986) Most radiation 

worker records had extensive uranium urinalysis from 1964 to 1989 that indicated intermittent and 

rare plutonium urinalysis, in comparison. 

Plutonium Bioassay

Since historical detection limits for W.R. Grace are not available, a review of detection limits for the 

1960s and 1970s at other AEC sites was made. The internal dosimetry section of the DOE Hanford 

site profile (ORAUT 2004a) reports an MDA of 0.05 dpm/sample. The Savannah River site profile 

(ORAUT 2005e) has a plutonium recording level and MDA of 0.1 dpm/1.5 L. These values are 

reasonably consistent with the observed reporting limits from W.R. Grace. From a review of bioassay 

records, the lowest observed nonzero recorded plutonium results at WR Grace was 0.03 dpm/L, 
Document No. ORAUT-TKBS-0043 Revision No. 00 Effective Date: 02/14/2006 Page 24 of 47

although some results are reported in units of dpm/sample, with a given sample volume. If additional 

information on detection capabilities is not available from the records, dose reconstructors should 

assume a decision level of 0.03 dpm/L and an MDA of 0.06 dpm/L.

Lung counting for 239Pu started at W.R. Grace in 1990. A germanium detector system had an ability to 

detect 239Pu but had a variable minimum detectable activity (MDA) of about 168 nCi (July 6, 1989) to 

481 nCi (October 22, 1987). The earlier Helgesson counting system likely could have been used for 

plutonium lung counting but was used only for uranium lung counting from about 1970. The detection 

limits are based on a cursory review of claimant files. Most results were non-detections or perhaps 

near the MDA. Urinalysis records are more useful for dose reconstruction purposes than lung 

counting because of the high MDAs of the lung counters.

Table 3-14. Organ dose for 1 yr for Hanford reference 10-yr-aged 

fuel-grade plutonium mixture (12%) and Hanford reference 5-yr-aged 

fuel-grade plutonium mixture (6%) for an annual chronic intake of 6.24 × 

105 pCi/yr and absorption type M.

a,b

Description

1-yr organ dose in rem for 

10-yr plutonium fuel grade

1-yr organ dose in rem for 

5-yr plutonium fuel grade

Adrenals 7.37E-03 5.63E-03

Bladder wall 7.37E-03 5.63E-03

Bone surface 3.66E+00 2.65E+00

Brain 7.37E-03 5.63E-03

Breast 7.37E-03 5.63E-03

Esophagus 7.37E-03 5.63E-03

ST wall 7.51E-03 5.73E-03

SI wall 7.58E-03 5.77E-03

ULI wall 8.71E-03 6.58E-03

LLI wall 1.11E-02 8.31E-03

Colon 9.76E-03 7.33E-03

Kidneys 8.10E-02 5.76E-02

Liver 7.70E-01 5.15E-01

Muscle 7.37E-03 5.63E-03

Ovaries 4.23E-02 2.91E-02

Pancreas 7.37E-03 5.63E-03

Red marrow 3.49E-01 2.57E-01

ET airways 1.84E+00 1.30E+00

Lungs 3.08E+00 2.14E+00

Skin 7.37E-03 5.63E-03

Spleen 7.37E-03 5.63E-03

Testes 4.28E-02 2.96E-02

Thymus 7.37E-03 5.63E-03

Thyroid 7.37E-03 5.63E-03

Uterus 7.37E-03 5.63E-03

Remainder 9.05E-03 6.82E-03

Effective dose 4.94E-01 3.46E-01

a. The table is for illustrative purposes only. It indicates the possible maximally exposed organs.

b. ICRP 2000, ICRP DCF database code.

Major plutonium processing sites that had ample air and bioassay sampling data can demonstrate the 

plutonium exposures that occurred in the DOE/AWE sites. In Attachment B of ORAUT (2005b), 

plutonium urine samples from Rocky Flats and Hanford for 1953 to 1969 were evaluated. Even 

though the median sample varied for most years, the value was less than 0.023 pCi/d. The Rocky 
Document No. ORAUT-TKBS-0043 Revision No. 00 Effective Date: 02/14/2006 Page 25 of 47

Flats/Hanford site analysis concluded that the assumption of 3,300 pCi/yr to which an individual could 

be exposed to type M or S characteristics is valid because these are the limiting absorption types for 

systemic organs. However, for non-systemic organs such as the lungs or respiratory tract for type S, 

this may not be the case. Table 3-15 lists all of the positive plutonium urinalysis results obtained from 

claimant files submitted by October 2005. 

Table 3-15. Plutonium urinalysis results for 

claims.

a

Sample numbera Date

Urinalysis in 

pCi/day

1 7/28/1968 0.08

2 8/23/1970 1.29

3b 10/3/1967 0.19

4b 10/11/1967 0.20

5b 11/10/1967 0.12

6b 11/19/1967 0.04

7b 12/13/1967 0.21

8b 3/3/1968 0.03

9b 3/29/1968 0.03

10 2/15/1973 0.02

11 12/5/1967 0.29

12 2/19/1969 0.02

13 12/5/1967 0.12

14 2/21/1969 0.16

15 12/13/1967 0.03

16 10/7/1971 0.10

17 10/16/1968 0.02

18 2/4/1967 0.06

19 5/20/1969 0.12

20 8/1/1969 0.04

21 12/7/1967 0.06

22 8/6/1968 0.44

23 8/6/1968 0.16

24 10/8/1967 0.04

25 10/15/1968 0.18

26 10/5/1967 0.03

27 11/3/1967 0.08

28 2/20/1969 0.04

29 12/7/1967 0.11

30 2/5/1968 0.73

31 8/5/1968 0.07

32 9/7/1967 0.10

33 11/10/1967 0.08

34 12/28/1967 0.06

35 12/5/1967 0.08

a. All data from claims filed as of 11/30/05.

b. These urinalysis results are from one claimant which 

also had the highest recorded excretion rate.
Document No. ORAUT-TKBS-0043 Revision No. 00 Effective Date: 02/14/2006 Page 26 of 47

3.3 THORIUM EXPOSURES

Thorium oxides, carbides, and hydroxides are absorption type S; nitrates and other compounds are 

type M. In the absence of specific information, the dose reconstructor should assume either type M or 

S whichever is more claimant-favorable. There is no air-sampling or urinalysis information available 

for the W.R. Grace facility for thorium. Therefore, either the gross alpha count data for early uranium 

air sampling from HASL reports 1959 and 1961, the urinalysis for uranium or the lung counts for 

uranium were used to estimate the annual intake. The most claimant favorable results will be used to 

estimate the thorium intake. Assuming equilibrium between 232Th and 228Th would be claimant 

favorable.

There were no dedicated facilities for thorium processing thorium operations utilized the same 

facilities as uranium operations. The low enriched uranium operations air sampling data is bounding 

for natural thorium. To account for unmonitored thorium exposures at the W.R. Grace/NFS plant it is 

assumed that the thorium intake is equal to the uranium intake by mass. Natural uranium has a lower 

specific activity than enriched uranium, so it is claimant favorable to assume natural uranium when 

determining the relative activity of thorium. To determine the relative activities of uranium to thorium, 

the specific activity of 232Th (1.1 × 10-7 Ci/g) is divided by the specific activity of natural uranium (6.83

× 10-7 Ci/g). This results in a relative 232Th-to-uranium intake fraction by activity of 0.161. Exposure 

from 226Ra (half-life of 5.75 years) is assumed to be insignificant because the thorium was likely to 

have been recently produced and because the dose conversion factor is small compared to thorium. 

Thorium oxide as ThO2 and uranium oxide as 233UO2 were the finished products for the production of 

233U/232/228Th MOX fuel. The 233U was received from Oak Ridge in the form of uranyl nitrate, then 

precipitated to (NH4)2-U2O7 (ADU), then calcined to UO2. Blending, pressing, and grinding were also a 

part of the process in addition to chemical processing. Particle sizes did vary. Because of criticality 

concerns, 4 kg of material could only be processed at one time. 

Usually tens of kilograms were processed per feasibility report or campaign. No enrichment was 

stated in the feasibility report, but the ZrO2-

233UO2 pellet feasibility report (Housholder 1963b) 

mentioned 93% enrichment. 

The combination of thorium and uranium oxides (232/228ThO2 and 233UO2) follows the following matrix: 

1% by weight 233UO2 / 99% by weight, 232/228ThO2 with a breakdown of 50% by activity 232Th and 50% 

by activity 228Th, and assuming equilibrium. It is not known if this material had aged substantially but 

232,/228Th/233U fuel can have considerable in-growth of many radionuclides because of the short half 

lives of their daughters.

There may have been several 232/228Th lung counts performed. These counts were likely below the 

MDAs of the time. A reasonable MDA for 232Th is 30 nCi. (ORNL 2002) 

Radon-220 from the decay of natural thorium is a potential source of dose to the lungs.

3.4 URANIUM AND PLUTONIUM DIOXIDE- MOX, MIXED OXIDE FUEL EXPOSURES

In this document, the proposed MOX fuel consists of 20% by weight of 3.5%-enriched 235UO2 and 

80% by weight of Hanford reference 10-yr-aged fuel-grade plutonium mixture (12%) PuO2. There is 

currently no specific process or operational information on MOX fuel recycling at W.R. Grace. 
Document No. ORAUT-TKBS-0043 Revision No. 00 Effective Date: 02/14/2006 Page 27 of 47

3.5 AMERICIUM

Americium is type M for all compounds. Note, the IMBA code can calculate type S intakes and thus 

the recommendation stated in section 3.2 of assigning 241Am as intake type S if the primary 

radionuclide is type S can be followed if deemed necessary. 

There may have been several 241Am lung counts performed. These counts were likely below the 

MDAs of the time. For example, a lung count performed in 1981 yielded a recorded result of 0.11 nCi 

but the reasonable MDA at the time was between 0.3 to 0.9 nCi. The current (2005) MDA would be 

about 0.1 nCi (ORAUT 2004f) 

3.6 OCCUPATIONAL INTERNAL DOSE RECONSTRUCTION SUMMARY AND 

ASSUMPTIONS - RESERVED

4.0

During operations at the facility, occupational exposure occurred from handling received material, 

standing near stored fissile material (either as feed or product), and airborne radioactivity with the 

resultant buildup of surface contamination. The fissile material was enriched UF6 arriving in approved 

shipping and storage cylinders, solid fissile material in various forms (liquid, powder, or metal) to be 

converted into or made into nuclear fuel, or scrap material containing LEU or HEU uranium.

ESTIMATION OF EXTERNAL EXPOSURE

4.1 RECORDED OCCUPATIONAL EXPOSURE AND EXTERNAL DOSIMETRY

Beginning in 1957, Chicago Nuclear supplied all dosimetry badges and performed the necessary 

calibrations. Landauer supplied all dosimetry beginning in 1961 using control film. The personnel 

monitoring reporting were normally in net exposure, and the control film reading was deducted from 

the personnel film reading. If the control film appeared to have been exposed differently from the 

personnel packets, the densities on the personnel film were normalized to Landauer controls only, 

and a non- control reading was reported. Minimal beta or soft X-ray skin dose readings were not 

reported until after a positive skin dose exposure was recorded (Koperski 2004). In 1989 Landauer 

supplied W.R. Grace/NFS with thermal luminescent dosimetry (TLD).

External radiation exposure recording at WR Grace/NFS utilized the U.S. AEC form AEC 5 similar to 

the NRC form 5 of today. Gamma, beta and neutron exposures are listed in separate columns. In 

1961, Landauer started to report the external dose in the same format as the AEC form 5. In 1964 

Landauer still reported the gamma, beta and neutron exposures in separate columns but had the 

same forms filled out separately for whole body, skin and extremities. Code 1 was for the whole body, 

code 2 for the skin of the whole body and code 3 for the hands and forearms. Up to December 31, 

1969, the reported skin dose represents only the shallow dose component (i.e., it does not include the 

penetrating gamma component). From January 1 1970, the reported skin exposure included the 

penetrating gamma component. For this reason, the shallow component should be calculated as the 

difference between the reported skin and total body dose. 

In addition, there was forearm and hand (extremity) monitoring at W.R. Grace/NFS from at least the 

1960s to the 1970s. The MDL information in Table 4-1 for wrist (forearm) and finger (hand or 

extremity) is from Landauer. Ring and wrist badges were calibrated only for high-energy gamma (for 

137Cs at 0.662 MeV) and high-energy beta (1.5 MeV) unless special arrangements were made with 

the Plant (Koperski 2004) MDL information for photon and beta whole body monitoring are also listed 
Document No. ORAUT-TKBS-0043 Revision No. 00 Effective Date: 02/14/2006 Page 28 of 47

in Table 4-1. The missed dose listed in Table 4-1 are considered default values and are to used if the 

available claimant dosimetry records are incomplete. The dosimetry used is further described in Table 

4-2.

Table 4-1. Minimum detectable level (MDL) and maximum potential missed photon or beta dose.

a

Dosimeter Period of use MDL (rem) Max annual missed dose (rem)

Nuclear Chicago film -

whole body

1957-5/1959b 0.01 photons

0.01 beta

0.26 beta -photons (every 2 weeks)

Nuclear Chicago film -

whole body

6/1959-12/1960b 0.01 photons

0.01 beta

0.12 beta -photons (monthly)

Landauer filmwhole body1/1961 – 12/1963b 0.01 photons

0.01 beta

0.12 beta -photons (monthly)

Landauer filmwhole body1/1964 – 12/1988c 0.01 photons

0.01 beta

0.12 beta -photons (monthly)

Landauer TLDwhole body1/1989 – 12/1998c 0.01 photons

0.01 beta

0.04 beta -photons (quarterly)

Landauer OSLwhole body1/1/1999-12/2004c 0.001 photons

0.001 beta

0.004 beta -photons (quarterly)

Landauer TLDwhole body1/2005 – presentc 0.01 photons

0.01 beta

0.04 beta -photons (quarterly)

Film-badgewrist1957- June 1974d 0.04 photons

0.04 beta

0.24 beta-photons (monthly)

Landauer Type M -

wrist

July 1974-About 1983c 0.02 photons 

0.04 beta

0.12 photons (monthly)

0.24 beta (monthly)

G5 wrist film badge –

wrist

About 1983 – 1990c 0.02 photons

0.04 beta

0.12 photons (monthly)

0.04 photons (quarterly)

0.24 beta (monthly)

0.08 beta (quarterly)

K5 TLD 

wrist 

1991- presentc 0.01 photons

0.04 beta

0.06 photons (monthly)

0.24 beta (monthly)

Film badgefinger1957-1982c 0.04 photons

0.04 beta

0.24 beta-photons (monthly)

U3 TLD (LiF) -

finger ring

About 1983 – presentc 0.03 photons

0.04 beta

0.18 photons (monthly)

0.08 photons (quarterly)

0.24 beta (monthly)

0.08 beta (quarterly)

a. TLD = thermoluminescent dosimeter.

b. MDLs from Tester (2005).

c. Koperski (2004). MDL information as communicated by Landauer Inc.

4.1.1

The NRC has reported on the annual occupational radiation exposure for nuclear power reactors 

since 1974. For facilities like W.R. Grace, consistent annual reporting could not be found in NRC 

records until 1982. The only information from before 1982 was for whole-body exposures in excess of 

1.25 rem for employees with employment of less than 90 d. The reporting format varied over the 

years. In the early years, only the number of monitored individuals and the measurable dose, 

collective dose, and average measurable doses in rem were provided. In later years, the annual 

whole-body doses were segregated into dose bins with the number of workers in each bin. Tables 4-3 

and 4-4 summarize the results of the research into NRC records.

NRC Annual Reports
Document No. ORAUT-TKBS-0043 Revision No. 00 Effective Date: 02/14/2006 Page 29 of 47

Table 4-2. Dosimetry for external whole-body, wrist, and extremity exposures. 

Period Monitoring technique Dosimeter description

Beta/photon dosimeters

1957 – June 1974

whole body

Photographic film badge Nuclear-Chicago or similar film badges. Nuclear-Chicago 

film badge contained single film packet. Three filters (front 

and back) were incorporated into film badge for energy 

dependence: cadmium, aluminum, and lead.

July 1974 – About 1983*

whole body

Landauer J (beta and 

gamma)

Type J dosimetry were film badges. Gamma and X-ray: 30 

keV to 20 MeV; beta: over 1.5 MeV.

About 1974* – April 1990

whole body

Landauer G1 Film emulsion packaged placed in standard Gardray 

holder/badge for monitoring beta, X-ray, and gamma 

exposure. Insensitive to neutron radiation. 

Gamma and X-ray: 30 keV to 20 MeV; beta: over 1.5 

MeV.

May 1990 – present

whole body

Landauer Z1 dosimeter Comprised of 3 TLD-700 chips for monitoring beta, X-ray 

and gamma exposure. Insensitive to neutron radiation.

Replaced Landauer G1.

July 1974 – About 1983*

wrist

Landauer Type M (wrist 

beta-gamma) badges.

Type M dosimetry was a film badge. Gamma and X-ray: 

30 keV to 20 MeV; beta: over 1.5 MeV.

About 1983* – 1990

wrist

G5 wrist film badge Responded to beta, X-ray, and gamma exposure to 

provide data on extremity dose. 

Gamma and X-ray: 30 keV to 20 MeV; beta: over 1.5 

MeV.

1991 – present

wrist

K5 TLD wrist badge Comprised of 3 TLD-100 chips.

1957-1974 Film badge-finger Nuclear-Chicago or similar film badges. Nuclear-Chicago 

film badge contained single film packet. Three filters (front 

and back) were incorporated into film badge for energy 

dependence: cadmium, aluminum, and lead.

About 1983* – present

finger ring

U3 TLD (LiF) Responded to beta, X-ray, and gamma exposure to 

provide data on extremity dose.

Gamma and X-ray: 30 keV to 20 MeV; beta: over 1.5 

MeV.

* This analysis found no documentation that shows the start of G1, U3, G5, I8, and E1 dosimeter use and the end of Type K, 

J, and M badges use.

Table 4-3. Annual occupational radiation exposures for 1982 to 1994. 

Year

Total workers 

monitored

Workers with 

measurable 

dose

Collective dose 

(person-rem)

Average 

measurable 

dose (rem) Reference

1982 1161 1014 59.77 0.06 NUREG-0714 Vol. 4 and 5a

1983 994 591 35.37 0.06

1984 904 626 37 0.06 NUREG-0713 Vol. 6b

1985 871 562 28 0.05 NUREG-0713 Vol. 7b

1986 996 389 22 0.06 NUREG-0713 Vol. 8b

1987 3559 367 20 0.05 NUREG-0713 Vol. 9b

1988 5516 627 32 0.05 NUREG-0713 Vol. 10b

1989 6136 698 40.525 0.06 NUREG-0713 Vol. 11b

1990 7373 739 46.025 0.06 NUREG-0713 Vol. 12b

1991 4347 830 41.750 0.05 NUREG-0713 Vol. 13b

1992 1,423 1,312 177.825 0.14 NUREG-0713 Vol. 14b

1993 3,373 323 78.650 0.24 NUREG-0713 Vol. 15b

1994 610 484 31.945 0.07 NUREG-0713 Vol. 16b

a. NRC (1985).
Document No. ORAUT-TKBS-0043 Revision No. 00 Effective Date: 02/14/2006 Page 30 of 47

b. NRC (1985–2003).
Document No. ORAUT-TKBS-0043 Revision No. 00 Effective Date: 02/14/2006 Page 31 of 47

Table 4-4. Annual occupational TEDE for 1995 to 2002.

a

Year

Number of Individuals with whole body doses in the ranges (rem)

Total no. 

monitored

No. 

with 

meas. 

dose

Total 

collective 

TEDE 

(rem)

Ave. 

meas. 

TEDE 

(rem)

NUREG-0713b

volume no.

No. 

meas. <0.1

0.1- 

0.25

0.25 -

0.5

0.5 – 

0.75

0.75 –

1.0

1.0 – 

2.0

2.0 – 

3.0

3.0 –

4.0

4.0 –

5.0

5.0 – 

12.0 >12.0

1995 192 166 12 16 8 0 0 0 0 0 0 0 392 200 15.185 0.08 Vol. 17

1996 178 195 26 26 7 3 0 0 0 0 0 0 435 257 24.937 0.097 Vol. 18

1997 180 323 38 27 8 0 4 0 0 0 0 0 578 396 33.656 0.065 Vol. 19

1998 126 387 66 45 16 13 7 0 0 0 0 0 660 534 64.951 0.122 Vol. 20

1999 58 559 68 38 15 2 0 0 0 0 0 0 740 682 46.227 0.068 Vol. 21

2000 155 377 62 55 15 4 8 0 0 0 0 0 676 521 60.703 0.177 Vol. 22

2001 70 498 104 64 21 10 12 2 0 0 0 0 781 711 95.599 0.134 Vol. 23

2002 92 628 159 92 20 8 11 3 0 0 0 0 1,013 921 115.848 0.126 Vol. 24

a. TEDE = total effective dose equivalent.

b. NRC (1985–2003).
Document No. ORAUT-TKBS-0043 Revision No. 00 Effective Date: 02/14/2006 Page 32 of 47

4.1.2

The majority of photons from natural uranium metals and solutions are in the 30 to 250 keV energy 

range. As uranium becomes more enriched or less enriched, the photon spectrum increases. Also the 

spectrum can become hardened since solid or liquid uranium objects provide considerable shielding. 

While it is recognized that solid or liquid uranium sources will have a hardened photon spectrum, 

exposure to a thin layer of uranium on a surface will result in a larger fraction of exposure to lower 

energy photons. (ORAUT-OTIB-0004 2005a) Considering the organ dose conversion factors and 

radiation effectiveness factors it would be claimant-favorable to assume that the workers were 

exposed to photon energies in the 30 to 250 keV range. 

Penetrating Dose Determination and Deep Dose Calculations

Thorium and its progeny create an increase in photon energies in the range of 30-250 keV; therefore 

it would be appropriate to assign energy that workers are exposed to in the 30-250 keV energy range. 

W. R. Grace/NFS distribution of low energy photons is dependent on the amounts, separation and 

enrichments of uranium and the age and type of plutonium used at the site. 

To ensure claimant favorability, a photon energy range of 100% 30 to 250 keV should be applied.

Table 4-2 lists a more detailed description of the Nuclear Chicago or Landauer dosimetry used at 

W.R. Grace/NFS.

4.1.3

Non-penetrating dose can be considered for this TBD as electrons greater than 15 keV and photons 

less than 30 keV if the employee worked at a plutonium facility. Actual assignment of the electron or 

low energy photon distribution is dependent upon the cancer location and radiation work environment 

of the claimant. The general approach to skin dose reconstruction is as follows: (ORAUT-OTIB-0017 

2005):

Non-penetrating Dose Determination and Skin Dose Calculations

1. Translate the reported doses into non-penetrating and penetrating doses.

2. Assign the penetrating dose as photons in the energy range of 30-250 keV.

3. Assign the non-penetrating dose as electrons > 15 keV (corrected to account for attenuation, if 

applicable), or photons < 30 keV if the employee worked in a plutonium facility.

4. Add missed electron and/or photon dose. 

5. Include missed neutron dose, if applicable.

6. All dose conversion factors for the skin should be assumed to be 1.

In general, non-penetrating radiation dose should be assigned as < 30 keV photons if the employee 

worked with or around plutonium. Otherwise, > 15 keV electrons should be assigned. If unknown, 

consider the following guidance:

1. For a likely non-compensable case, it is acceptable to assume the non-penetrating dose is 

associated with < 30 keV photons, as this is claimant favorable. 

2. For a likely compensable case, it is acceptable to assume the non-penetrating dose is 

associated with > 15 keV electrons. 
Document No. ORAUT-TKBS-0043 Revision No. 00 Effective Date: 02/14/2006 Page 33 of 47

3. If the compensability decision may hinge on this issue, and if the partitioning of the nonpenetrating dose cannot be based on the available information, additional research may be 

required. 

As noted earlier, prior to 1969, the reported skin dose represents only the shallow dose component 

and after 1969 the reported skin exposure includes the penetrating gamma component. For this 

reason, the shallow component should be calculated as the difference between the reported skin and 

total body dose. 

For film dosimetry, there is an over-response to 16 keV x-rays and 59 keV photons by factors of 8.5 to 

12 and 14 to 19 respectively. (Wilson, 1990) This would have bound response to lower energy 

uranium photons as well. The general approach to skin dose calculation for W.R. Grace/NFS is as 

follows: 

Measured dose 

1. Subtract the reported whole body or total body (deep) dose from the skin of the whole body or 

skin of the total body dose starting in 1969. This will result in giving the nonpenetrating/shallow dose component. Before1969, the skin dose was assigned only as the 

measured beta component. 

2. Assign the calculated non-penetrating dose as electrons > 15 keV. A correction factor should 

be provided for clothing, if applicable, depending on likely clothing thickness and beta energy. 

Assuming two pairs of coveralls including the paper lining (0.80) and one Dacron cotton lab 

coat (0.91) the midpoint correction factor would be about 0.86 for uranium betas. Use Table 6-

11 on page 6-23 of DOE 2000.

3. Assign the reported whole body or deep dose to the energies in the 30 to 250 keV range. 

4. For Chemical/Production Operators only assign the calculated neutron dose partitioned in the 

energy range of 0.1 to 2 MeV and use an organ DCF of 1. 

Missed Dose

1. For any badge cycle with a zero or minimal result in either the whole body or beta reading or 

both, assign a single missed dose as explained in items 6-8 below.

2. If only the whole body or deep dose was reported as zero or minimal (designated with a “M” 

(minimal) in the records for results at or below the minimal detectable level of 10 mrem), the 

missed dose assigned should be the appropriate MDL (gamma/x-ray) for that era (divided by 

2, treated as lognormal with a GSD of 1.52) and considered to be photons in the energy range 

form 30 to 250 keV Use Table 4-1.

3. For any badge cycle with a zero result in either the skin or beta recorded dose, the missed 

dose can be assigned based on the appropriate MDL of the beta of that era (divided by 2, 

treated as lognormal) and considered to be electrons corrected by a factor 0.86 or low energy 

photons. Use Table 4-1.

4. If both the whole body or deep dose and the skin or beta readings were reported as zero, the 

missed dose assigned should be the appropriate MDL (gamma/x-ray) for that era (divided by 

2, treated as lognormal) and considered to be 30-250 keV photons. 
Document No. ORAUT-TKBS-0043 Revision No. 00 Effective Date: 02/14/2006 Page 34 of 47

5. For Chemical/Production Operators only assign neutron dose according to section 4.1.4, 

Neutron dose.

4.1.4

In order to estimate neutron dose to unmonitored workers a 0.20 to 1 neutron to photon ratio based 

upon their photon exposure is recommended. This should be applied to “Chemical/Production

Operators” for which dosimetry information is available or based upon their missed dose photon 

calculation. See also Table 4-16. 

Neutron Dose Assessment

4.2 ENRICHED URANIUM SOURCE TERM INFORMATION

ORAUT (2004c) assessed the potential deep dose to a worker from fissile material in a 5-by-5 

double-stacked birdcage array for 2,000 hr/yr used the same methods for nuclear criticality control 

(geometry and mass control). The calculation showed a dose result of 0.13 R/yr for a worker 1 m from 

the birdcage array and the 95th-percentile exposure of 0.76 R/yr for a worker 1 foot from the array. 

Table 4-5 lists the annual organ doses due to the potential exposure to the stacked birdcage array.

4.3 SURFACE CONTAMINATION 

Facility operations created areas of surface contamination at W.R. Grace. A component of external 

exposure is from the settling of airborne radioactivity particles onto surfaces in the work area as 

described in the guidance for estimating the maximum plausible dose at atomic weapons employer 

Table 4-5. Annual organ doses due to exposure to a 5-by-5 stacked array of 

birdcages. 

Organ

Annual organ dose (rem)

Photons E=30-250 keV Photons E>250 keV Total

Bladder 8.09E-02 5.74E-02 1.38E-01

Red bone marrow 4.07E-02 4.68E-02 8.75E-02

Bone surface 7.99E-02 4.97E-02 1.30E-01

Breast 8.23E-02 6.05E-02 1.43E-01

Colon 6.89E-02 5.49E-02 1.24E-01

Esophagus 4.47E-02 4.84E-02 9.31E-02

Eye 8.03E-02 5.72E-02 1.38E-01

Ovaries 6.21E-02 5.32E-02 1.15E-01

Testes 9.32E-02 6.12E-02 1.54E-01

Liver 6.92E-02 5.49E-02 1.24E-01

Lung 6.41E-02 5.47E-02 1.19E-01

Remainder organs 5.71E-02 5.12E-02 1.08E-01

Skin 5.80E-02 5.43E-02 1.12E-01

Stomach 8.13E-02 5.75E-02 1.39E-01

Thymus 9.15E-02 5.80E-02 1.50E-01

Thyroid 9.36E-02 6.32E-02 1.57E-01

Uterus 6.57E-02 5.11E-02 1.17E-01

Source: ORAUT (2004c).

facilities (ORAUT 2005a). The settling of radioactive particles on surfaces establishes an area-wide 

exposure source term and an external exposure pathway. Dose coefficients by radionuclide for 

assessing the external exposure to organs from surface contamination (i.e., a ground plane) are 

provided in Federal Guidance Report No. 13 (EPA 2002). 
Document No. ORAUT-TKBS-0043 Revision No. 00 Effective Date: 02/14/2006 Page 35 of 47

The available documents are not clear as to the resultant surface contamination levels in each of the 

buildings where radioactive material operations occurred during the time the site was an AWE facility.

Therefore, to provide the component of external exposure for dose estimation, the surface 

contamination levels were assumed at certain values that should encompass surface contamination 

levels found at this facility.

Four surface contamination levels were selected for dose estimation. The lowest surface 

contamination level is the facility’s administrative control limit of 5,000 dpm/100 cm2

(2.25E+07 pCi/m2

), above which surface contamination cleanup would be required (AEC 1959, 1961; 

Congress 1986). This level should be applied to building areas (office spaces, cafeteria, etc.) that 

would not have processed radioactive material but could have been contaminated by the nearby 

process operations. The exposure period should encompass the entire work year (i.e., 2,000 hr).

The second surface contamination level is based on the guidance for AWE facilities for the maximum 

air concentration from 365 d of constant disposition (ORAUT 2005a). The contamination level for this 

case would be approximately 8.1 E+07 pCi/m2

. This level is higher than the maximum surface 

contamination levels provided in the two AEC survey documents (AEC 1959, 1961). Because this 

level of contamination is only for process areas, the time of exposure is the annual work time minus 

1 hr for lunch each workday.

The last two surface contamination cases consider the highest stated contamination level found in any 

publicly available document for this facility. Based on worker testimony given to the U.S. Congress, 

the highest value stated is approximately 100,000 dpm/100 cm2 with an average of approximately 

45,000 dpm/100 cm2 from all of the values in Congress (1986). These values translate into 

approximately 4.50E+08 pCi/m2 and 2.00E+08 pCi/m2

, respectively. Again, the time of exposure is 

only when a worker is in a process area, which is the annual work time minus 1 hr for lunch each 

workday.

The annual organ dose due to exposure to contaminated surfaces was estimated by multiplying the 

surface contamination level by the dose coefficients for contaminated ground surfaces for the 

applicable radionuclides. Using ground contamination dose conversion coefficients would be claimant 

favorable when used for surface contamination near the organ of interest. This is because the 

conversion factors for the ground would have an inverse square component built into them. So, for 

example, dose to the stomach would be different for equivalent contamination levels if on the ground 

versus a work surface near the abdomen. Because the facility operations varied by building, the 

annual organ dose estimates were determined for the combination of the radionuclides for a specific 

operation. Based on the process descriptions, seven radionuclide groups were assessed for external 

exposure, thorium, 233U/thorium, 93%-enriched uranium, 3.5%-enriched uranium, DU, plutonium, and 

MOX fuel.

For the 233U/thorium annual organ doses, the percentage of 234U and thorium in the surface 

contamination is necessary to determine the total annual dose. Documentation on the manufacture of 

233U/thorium fuel for Shippingport is not available. However, historical information implies that the fuel 

mixture would be 20% 233U and 80% ThO2 by weight (WNA 2005). This mixture was applied to the 

four contamination levels. The fraction of specific activity for thorium isotopes is associated with 230Th

and 232Th. Based on a ORAUT (2005c), weight fractions of 0.116 for 230Th and 0.884 for 232Th are 

applied in this assessment. 

To assess the annual external organ dose from uranium operations, the isotopic fractions for HEU, 

EU, and DU must be set. Based on information for Hanford (ORAUT 2004a), Table 4-6 shows the 

activity fractions of these forms of uranium, which were applied for the annual organ doses.
Document No. ORAUT-TKBS-0043 Revision No. 00 Effective Date: 02/14/2006 Page 36 of 47

Table 4-6. Uranium activity fractions. 

Uranium form

Activity fractions

234U 235U 236U 238U

93% enricheda 0.968 0.030 0.002 0.0003

3.5% enriched 0.818 0.034 --- 0.147

DU 0.155 0.011 0.0005 0.834

Source: ORAUT (2005d).

a. These values also correspond to the specific activities 

provided in Table C.1 of ANSI 1995.

For the production of MOX fuel for SEFOR, the mixture was 20% reactor-grade PuO2 and 

80%-depleted UO2 (Sharma 2002). This same mixture was assumed for the W.R. Grace facility during 

the times of MOX fuel production. The isotopic activity fractions for DU were previously presented. For 

the isotopic activity fractions for plutonium radionuclides, the time since separation must be known 

because 241Am will build up in the PuO2. It was assumed that the reactor-grade plutonium used to 

produce MOX fuel by the facility was not kept very long in storage before being mixed with DU for the 

MOX fuel. For this analysis, a storage time of 5 yr was assumed with the isotopic activity fractions 

from DOE Standard 1128 (DOE 2005). Therefore, the isotopic activity fractions for 5-yr-aged 

reactor-grade plutonium would be 0.0064 for 238Pu, 0.0206 for 239Pu, 0.0107 for 240Pu, 0.9537 for 

241Pu, and 0.0086 for 241Am.

Tables 4-7 to 4-13 show the annual organ dose estimates for the radionuclide combinations.

Table 4-7. External exposure from thorium surface contamination. 

Organ

Annual dose for 

2.25E+07 pCi/m2 (rem)

Annual dose for 

7.5E+07 pCi/m2 (rem)

Annual dose for 

2.0E+08 pCi/m2 (rem)

Annual dose for 

4.5E+08 pCi/m2 (rem)

R_marrow 1.37E-04 3.99E-04 1.22E-03 2.73E-03

Adrenals 9.24E-05 2.69E-04 8.21E-04 1.85E-03

B_surface 6.76E-04 1.97E-03 6.01E-03 1.35E-02

Brain 9.10E-05 2.65E-04 8.09E-04 1.82E-03

Breasts 6.86E-04 2.00E-03 6.10E-03 1.37E-02

GB_wall 8.97E-05 2.61E-04 7.97E-04 1.79E-03

Esophagus 7.45E-05 2.17E-04 6.62E-04 1.49E-03

ST_wall 1.07E-04 3.13E-04 9.54E-04 2.15E-03

SI_wall 8.78E-05 2.56E-04 7.81E-04 1.76E-03

ULI_wall 9.31E-05 2.72E-04 8.27E-04 1.86E-03

LLI_wall 9.24E-05 2.69E-04 8.21E-04 1.85E-03

HT_wall 1.02E-04 2.96E-04 9.03E-04 2.03E-03

Kidneys 1.21E-04 3.53E-04 1.08E-03 2.42E-03

Liver 1.06E-04 3.10E-04 9.45E-04 2.13E-03

Lung 1.17E-04 3.40E-04 1.04E-03 2.33E-03

Ovaries 1.09E-04 3.19E-04 9.71E-04 2.19E-03

Pancreas 8.33E-05 2.43E-04 7.41E-04 1.67E-03

Skin 4.13E-03 1.21E-02 3.67E-02 8.27E-02

Spleen 1.03E-04 3.02E-04 9.19E-04 2.07E-03

Testes 5.22E-04 1.52E-03 4.64E-03 1.04E-02

Thymus 1.20E-04 3.50E-04 1.07E-03 2.40E-03

Thyroid 1.73E-04 5.05E-04 1.54E-03 3.46E-03

UB_wall 1.08E-04 3.16E-04 9.64E-04 2.17E-03

Uterus 8.72E-05 2.54E-04 7.75E-04 1.74E-03

Muscle 4.59E-04 1.34E-03 4.08E-03 9.18E-03

H_remaind 4.26E-04 1.24E-03 3.79E-03 8.53E-03

H_E 2.85E-04 8.32E-04 2.54E-03 5.71E-03
Document No. ORAUT-TKBS-0043 Revision No. 00 Effective Date: 02/14/2006 Page 37 of 47

Table 4-8. External exposure from 233U/thorium surface contamination. 

Organ

Annual dose for 

2.25E+07 pCi/m2 (rem)

Annual dose for 

7.5E+07 pCi/m2 (rem)

Annual dose for 

2.0E+08 pCi/m2 (rem)

Annual dose for 

4.5E+08 pCi/m2 (rem)

R_marrow 1.53E-04 4.45E-04 1.36E-03 3.05E-03

Adrena 1.06E-04 3.08E-04 9.39E-04 2.11E-03

B_surface 7.00E-04 2.04E-03 6.23E-03 1.40E-02

Brain 1.04E-04 3.05E-04 9.29E-04 2.09E-03

Breasts 7.02E-04 2.05E-03 6.24E-03 1.40E-02

GB_wall 1.03E-04 3.00E-04 9.13E-04 2.06E-03

Esophagus 8.67E-05 2.53E-04 7.71E-04 1.73E-03

ST_wall 1.22E-04 3.55E-04 1.08E-03 2.43E-03

SI_wall 1.01E-04 2.94E-04 8.96E-04 2.02E-03

ULI_wall 1.07E-04 3.11E-04 9.48E-04 2.13E-03

LLI_wall 1.06E-04 3.08E-04 9.40E-04 2.12E-03

HT_wall 1.15E-04 3.36E-04 1.02E-03 2.31E-03

Kidneys 1.37E-04 3.99E-04 1.22E-03 2.74E-03

Liver 1.21E-04 3.52E-04 1.07E-03 2.41E-03

Lung 1.32E-04 3.84E-04 1.17E-03 2.63E-03

Ovaries 1.22E-04 3.57E-04 1.09E-03 2.45E-03

Pancreas 9.57E-05 2.79E-04 8.50E-04 1.91E-03

Skin 4.02E-03 1.17E-02 3.57E-02 8.04E-02

Spleen 1.17E-04 3.43E-04 1.04E-03 2.35E-03

Testes 5.42E-04 1.58E-03 4.82E-03 1.08E-02

Thymus 1.34E-04 3.91E-04 1.19E-03 2.68E-03

Thyroid 1.91E-04 5.58E-04 1.70E-03 3.83E-03

UB_wall 1.23E-04 3.59E-04 1.10E-03 2.46E-03

Uterus 1.00E-04 2.93E-04 8.93E-04 2.01E-03

Muscle 4.74E-04 1.38E-03 4.21E-03 9.48E-03

H_remaind 4.41E-04 1.29E-03 3.92E-03 8.83E-03

H_E 3.00E-04 8.76E-04 2.67E-03 6.00E-03

Table 4-9. External exposure from 93%-enriched uranium surface contamination. 

Organ

Annual dose for 

2.25E+07 pCi/m2 (rem)

Annual dose for 

7.5E+07 pCi/m2 (rem)

Annual dose for 

2.0E+08 pCi/m2 (rem)

Annual dose for 

4.5E+08 pCi/m2 (rem)

R_marrow 2.57E-03 7.51E-03 2.29E-02 5.15E-02

Adrenals 2.26E-03 6.58E-03 2.00E-02 4.51E-02

B_surface 6.35E-03 1.85E-02 5.65E-02 1.27E-01

Brain 2.32E-03 6.76E-03 2.06E-02 4.64E-02

Breasts 3.67E-03 1.07E-02 3.26E-02 7.34E-02

GB_wall 2.23E-03 6.51E-03 1.98E-02 4.46E-02

Esophagus 2.06E-03 6.00E-03 1.83E-02 4.11E-02

ST_wall 2.43E-03 7.09E-03 2.16E-02 4.86E-02

SI_wall 2.26E-03 6.60E-03 2.01E-02 4.53E-02

ULI_wall 2.34E-03 6.82E-03 2.08E-02 4.68E-02

LLI_wall 2.34E-03 6.82E-03 2.08E-02 4.68E-02

Twill 2.34E-03 6.81E-03 2.08E-02 4.67E-02

Kidneys 2.44E-03 7.12E-03 2.17E-02 4.88E-02

Liver 2.43E-03 7.08E-03 2.16E-02 4.86E-02

Lung 2.56E-03 7.47E-03 2.28E-02 5.13E-02

Ovaries 2.29E-03 6.69E-03 2.04E-02 4.59E-02

Pancreas 2.19E-03 6.38E-03 1.94E-02 4.37E-02

Skin 8.77E-03 2.56E-02 7.80E-02 1.75E-01

Spleen 2.44E-03 7.11E-03 2.17E-02 4.88E-02

Testes 3.48E-03 1.01E-02 3.09E-02 6.96E-02

Thymus 2.34E-03 6.84E-03 2.08E-02 4.69E-02

Thyroid 2.69E-03 7.83E-03 2.39E-02 5.37E-02

UB_wall 2.42E-03 7.05E-03 2.15E-02 4.83E-02

Uterus 2.28E-03 6.65E-03 2.03E-02 4.56E-02

Muscle 3.30E-03 9.61E-03 2.93E-02 6.59E-02

H_remaind 3.21E-03 9.37E-03 2.85E-02 6.42E-02

H_E 2.86E-03 8.34E-03 2.54E-02 5.72E-02
Document No. ORAUT-TKBS-0043 Revision No. 00 Effective Date: 02/14/2006 Page 38 of 47

Table 4-10. External exposure from 3.5%-enriched uranium surface contamination. 

Organ

Annual dose for 

2.25E+07 pCi/m2 (rem)

Annual dose for 

7.5E+07 pCi/m2 (rem)

Annual dose for 

2.0E+08 pCi/m2 (rem)

Annual dose for 

4.5E+08 pCi/m2 (rem)

R_marrow 2.89E-03 8.43E-03 2.57E-02 5.78E-02

Adrenals 2.54E-03 7.41E-03 2.26E-02 5.08E-02

B_surface 7.07E-03 2.06E-02 6.28E-02 1.41E-01

Brain 2.61E-03 7.63E-03 2.32E-02 5.23E-02

Breasts 3.99E-03 1.17E-02 3.55E-02 7.99E-02

GB_wall 2.52E-03 7.34E-03 2.24E-02 5.03E-02

Esophagus 2.32E-03 6.77E-03 2.06E-02 4.64E-02

ST_wall 2.74E-03 7.99E-03 2.43E-02 5.48E-02

SI_wall 2.55E-03 7.44E-03 2.27E-02 5.10E-02

ULI_wall 2.64E-03 7.69E-03 2.34E-02 5.27E-02

LLI_wall 2.64E-03 7.69E-03 2.34E-02 5.27E-02

HT_wall 2.63E-03 7.67E-03 2.34E-02 5.26E-02

Kidneys 2.74E-03 8.01E-03 2.44E-02 5.49E-02

Liver 2.73E-03 7.98E-03 2.43E-02 5.47E-02

Lung 2.89E-03 8.42E-03 2.57E-02 5.77E-02

Ovaries 2.58E-03 7.53E-03 2.30E-02 5.17E-02

Pancreas 2.47E-03 7.19E-03 2.19E-02 4.93E-02

Skin 9.06E-03 2.64E-02 8.06E-02 1.81E-01

Spleen 2.75E-03 8.02E-03 2.44E-02 5.50E-02

Testes 3.82E-03 1.11E-02 3.39E-02 7.63E-02

Thymus 2.64E-03 7.69E-03 2.34E-02 5.27E-02

Thyroid 3.01E-03 8.77E-03 2.67E-02 6.02E-02

UB_wall 2.72E-03 7.94E-03 2.42E-02 5.44E-02

Uterus 2.57E-03 7.50E-03 2.29E-02 5.14E-02

Muscle 3.63E-03 1.06E-02 3.23E-02 7.26E-02

H_remaind 3.54E-03 1.03E-02 3.15E-02 7.09E-02

H_E 3.18E-03 9.27E-03 2.82E-02 6.35E-02

Table 4-11. External exposure from DU surface contamination. 

Organ

Annual dose for 

2.25E+07 pCi/m2 (rem)

Annual dose for 

7.5E+07 pCi/m2 (rem)

Annual dose for 

2.0E+08 pCi/m2 (rem)

Annual dose for 

4.5E+08 pCi/m2 (rem)

R_marrow 9.81E-04 2.86E-03 8.72E-03 1.96E-02

Adrena 8.34E-04 2.43E-03 7.41E-03 1.67E-02

B_surface 2.59E-03 7.55E-03 2.30E-02 5.18E-02

Brain 8.51E-04 2.48E-03 7.57E-03 1.70E-02

Breasts 1.79E-03 5.23E-03 1.59E-02 3.59E-02

GB wall 8.21E-04 2.39E-03 7.30E-03 1.64E-02

Esophagus 7.53E-04 2.20E-03 6.70E-03 1.51E-02

ST_wall 9.02E-04 2.63E-03 8.02E-03 1.80E-02

SI_wall 8.31E-04 2.42E-03 7.39E-03 1.66E-02

ULI_wall 8.60E-04 2.51E-03 7.64E-03 1.72E-02

LLI_wall 8.60E-04 2.51E-03 7.64E-03 1.72E-02

HT_wall 8.65E-04 2.52E-03 7.69E-03 1.73E-02

Kidneys 9.20E-04 2.68E-03 8.18E-03 1.84E-02

Liver 8.99E-04 2.62E-03 7.99E-03 1.80E-02

Lung 9.51E-04 2.77E-03 8.45E-03 1.90E-02

Ovaries 8.59E-04 2.50E-03 7.63E-03 1.72E-02

Pancreas 8.02E-04 2.34E-03 7.13E-03 1.60E-02

Skin 5.83E-03 1.70E-02 5.19E-02 1.17E-01

Spleen 8.99E-04 2.62E-03 7.99E-03 1.80E-02

Testes 1.61E-03 4.69E-03 1.43E-02 3.21E-02

Thymus 8.81E-04 2.57E-03 7.83E-03 1.76E-02

Thyroid 1.05E-03 3.06E-03 9.32E-03 2.10E-02

UB_wall 8.98E-04 2.62E-03 7.99E-03 1.80E-02

Uterus 8.37E-04 2.44E-03 7.44E-03 1.67E-02

Muscle 1.48E-03 4.31E-03 1.31E-02 2.95E-02

H_remaind 1.42E-03 4.15E-03 1.26E-02 2.84E-02

H_E 1.19E-03 3.47E-03 1.06E-02 2.38E-02
Document No. ORAUT-TKBS-0043 Revision No. 00 Effective Date: 02/14/2006 Page 39 of 47

Table 4-12. External exposure from plutonium surface contamination. 

Organ

Annual dose for 

2.25E+07 pCi/m2 (rem)

Annual dose for 

7.5E+07 pCi/m2 (rem)

Annual dose for 

2.0E+08 pCi/m2 (rem)

Annual dose for 

4.5E+08 pCi/m2 (rem)

R_marrow 9.08E-05 2.65E-04 8.07E-04 1.82E-03

Adrenals 7.77E-05 2.27E-04 6.90E-04 1.55E-03

B_surface 5.02E-04 1.46E-03 4.46E-03 1.00E-02

Brain 8.15E-05 2.38E-04 7.24E-04 1.63E-03

Breasts 2.12E-04 6.18E-04 1.88E-03 4.23E-03

GB_wall 7.64E-05 2.23E-04 6.79E-04 1.53E-03

Esophagus 6.26E-05 1.83E-04 5.57E-04 1.25E-03

ST_wall 9.52E-05 2.78E-04 8.46E-04 1.90E-03

SI_wall 7.73E-05 2.26E-04 6.87E-04 1.55E-03

ULL_wall 8.31E-05 2.42E-04 7.39E-04 1.66E-03

LLI_wall 8.05E-05 2.35E-04 7.16E-04 1.61E-03

HT_wall 8.83E-05 2.57E-04 7.85E-04 1.77E-03

Kidneys 1.02E-04 2.98E-04 9.07E-04 2.04E-03

Liver 9.56E-05 2.79E-04 8.50E-04 1.91E-03

Lung 1.06E-04 3.10E-04 9.44E-04 2.12E-03

Ovaries 7.89E-05 2.30E-04 7.01E-04 1.58E-03

Pancreas 7.25E-05 2.12E-04 6.45E-04 1.45E-03

Skin 5.74E-04 1.67E-03 5.10E-03 1.15E-02

Spleen 9.53E-05 2.78E-04 8.47E-04 1.91E-03

Testes 1.91E-04 5.58E-04 1.70E-03 3.83E-03

Thymus 1.04E-04 3.04E-04 9.27E-04 2.09E-03

Thyroid 1.20E-04 3.51E-04 1.07E-03 2.41E-03

UB_wall 9.39E-05 2.74E-04 8.34E-04 1.88E-03

Uterus 7.53E-05 2.19E-04 6.69E-04 1.51E-03

Muscle 1.64E-04 4.78E-04 1.46E-03 3.28E-03

H_remaind 1.57E-04 4.57E-04 1.39E-03 3.14E-03

H_E 1.31E-04 3.82E-04 1.16E-03 2.62E-03

Table 4-13. External exposure from MOX surface contamination. 

Organ

Annual dose for 

2.25E+07 pCi/m2 (rem)

Annual dose for 

7.5E+07 pCi/m2 (rem)

Annual dose for 

2.0E+08 pCi/m2 (rem)

Annual dose for 

4.5E+08 pCi/m2 (rem)

R_marrow 8.03E-04 2.34E-03 7.14E-03 1.61E-02

Adrenals 6.82E-04 1.99E-03 6.07E-03 1.36E-02

B_surface 2.17E-03 6.33E-03 1.93E-02 4.34E-02

Brain 6.97E-04 2.03E-03 6.20E-03 1.39E-02

Breasts 1.48E-03 4.31E-03 1.31E-02 2.95E-02

GB_wall 6.72E-04 1.96E-03 5.97E-03 1.34E-02

Esophagus 6.15E-04 1.79E-03 5.47E-03 1.23E-02

ST_wall 7.41E-04 2.16E-03 6.59E-03 1.48E-02

SI_wall 6.80E-04 1.98E-03 6.05E-03 1.36E-02

ULI_wall 7.04E-04 2.05E-03 6.26E-03 1.41E-02

LLI_wall 7.04E-04 2.05E-03 6.26E-03 1.41E-02

HT_wall 7.09E-04 2.07E-03 6.31E-03 1.42E-02

Kidneys 7.56E-04 2.21E-03 6.72E-03 1.51E-02

Liver 7.38E-04 2.15E-03 6.56E-03 1.48E-02

Lung 7.82E-04 2.28E-03 6.95E-03 1.56E-02

Ovaries 7.03E-04 2.05E-03 6.25E-03 1.41E-02

Pancreas 6.56E-04 1.91E-03 5.83E-03 1.31E-02

Skin 4.78E-03 1.39E-02 4.25E-02 9.57E-02

Spleen 7.38E-04 2.15E-03 6.56E-03 1.48E-02

Testes 1.32E-03 3.86E-03 1.18E-02 2.65E-02

Thymus 7.25E-04 2.12E-03 6.45E-03 1.45E-02

Thyroid 8.63E-04 2.52E-03 7.67E-03 1.73E-02

UB_wall 7.38E-04 2.15E-03 6.56E-03 1.48E-02

Uterus 6.85E-04 2.00E-03 6.09E-03 1.37E-02

Muscle 1.21E-03 3.54E-03 1.08E-02 2.43E-02

H_remainder 1.17E-03 3.41E-03 1.04E-02 2.34E-02
Document No. ORAUT-TKBS-0043 Revision No. 00 Effective Date: 02/14/2006 Page 40 of 47

4.4 PLUTONIUM OPERATIONS

Due to the lack of information about plutonium operations at the W.R. Grace facility, certain 

assumptions were necessary for estimating the external exposure from MOX fuel production. The 

assumptions relate to the classification of the PuO2, the buildup of 241Am, and the extent of 

contamination outside of gloveboxes.

The classification or grade of plutonium is based on the source of the material because this sets the 

isotopic content of the PuO2. In general, there are two classifications of plutonium, namely weapons 

grade and reactor grade (DOE 2005). Weapons-grade plutonium has a relatively short reactor 

exposure time. Reactor-grade plutonium is from reprocessing DOE spent nuclear fuel or from 

research reactors. For radiological exposures, reactor-grade plutonium has higher radiation levels 

than weapons-grade plutonium due to the buildup of 241Am from the beta decay of 241Pu (Carbaugh 

2003; DOE 2005). Therefore, it was assumed that reactor-grade plutonium was used in the production 

of MOX fuel at the facility.

A storage time for buildup of 241Am was assumed based on the isotopic content in Table 8.4 of 

Carbaugh (2003); it is reasonable to assume that the storage time for PuO2 at the facility would be no 

greater than 5 yr. Table 4-14 lists the isotopic content and specific activity of plutonium based on 

these assumptions.

Table 4-14. Plutonium isotopic content and specific activity. 

Isotope

Decay 

modea Contenta(wt%)

Specific activitya

(Ci/g)

Contentb

(wt%)

Specific activity after 

5 yr of agingb (Ci/g)

236Pu Alpha 3.0E-06 53.4 NA NA

238Pu Alpha 0.58 17.1 0.10 1.64E-02

239Pu Alpha 72.10 6.22E-02 84.8 5.26E-02

240Pu Alpha 19.15 0.229 12.0 2.72E-02

241Pu Alpha

Beta

6.29 2.52E-03

103

3.0 2.43E+00 

NA

242Pu Alpha 1.88 3.93E-03 0.1 3.93E-06

241Am Alpha 0.02 3.43 0.0 2.19E-02

a. DOE 2005.

b. Carbaugh 2003.

Section 6.2.2 of DOE (2005) provided results for measured extremity photon dose rates from 

plutonium glovebox operations . Two glovebox cases with reactor-grade plutonium based on the 

isotopic concentration are shown in the third column of Table 4-14. 

As discussed on pages 6-7 of DOE 2005:

Doses to the extremities are usually dominated by gamma rays in typical glovebox operations. 

[...] The extremity dose is more limiting than a whole body dose if the dose gradient is greater 

than 10:1 over a distance of 1 meter, the maximum distance from the fingers to the trunk of the 

body. In most cases, the source is not at arm’s length and the dose gradient needs to be 10:1 

or 20:1 for the extremity dose to be limiting (NUREG/CR-4297; Reece et al. 1985). But in 

highly shielded gloveboxes, it is possible to have a very high extremity dose from dust layer on 

gloves; the dose to the torso can be much lower because of shielding applied to the glovebox.
Document No. ORAUT-TKBS-0043 Revision No. 00 Effective Date: 02/14/2006 Page 41 of 47

4.5 MEDICAL X-RAYS

Site workers, whether under W.R. Grace or NFS, received an annual occupationally related diagnostic 

medial X-ray and a pre-employment X-ray. The exposure geometry was posterior-anterior. The air 

kerma at skin entrance for the diagnostic chest X-ray was estimated to be 0.2 R for pre-1970 X-rays, 

0.1 R for 1970 to 1985, and 0.05 R for 1986 to the present. To date no site-specific information is 

available for W.R. Grace. Photofluorography (PFG) could also have been possible but unless there is 

evidence presented that PFG was performed in the claimant files it is reasonable to assume that 

PFGs were not performed. Table 4-19 shows the annual organ doses due to the assumed annual 

diagnostic chest X-ray (ORAUT 2005d). The values in Table 4-15 should be entered into the 

Interactive RadioEpidemiology Program (IREP) software as the annual dose to an acute exposure to 

photons with energy from 30 to 250 keV. The distribution is assumed to be normal with a standard 

deviation of 30%. 

4-15. Annual organ doses due to the assumed annual diagnostic chest X-ray. 

Organ Annual dose (rem) Annual dose (rem) Annual dose (rem)

Period 1957-1969 1970-1985 1985-present

Bladder 0.025 0.0001 0.00026

Red bone marrow 0.0184 m

0.0172 f

0.0092 0.0089

Bone surface 0.0902 0.0451 0.0337

Breast 0.0098 0.0049 0.0058

Colon/rectum 0.025 0.0001 0.00026

Esophagus 0.0902 0.0451 0.0337

Eye 0.00640 0.0032 0.0039

Ovaries 0.025 0.0001 0.00026

Testes 0.005 0.000001 0.0000005

Liver/gall bladder/spleen 0.0902 0.0451 0.0337

Lung 0.0838 m

0.0902 f

0.0451 0.0337

Remainder organs 0.0902 0.0451 0.0337

Skin 0.270 0.135 0.070

Stomach 0.0902 0.0451 0.0337

Thymus 0.0902 0.0451 0.0337

Thyroid 0.0348 0.0032 0.0039

Uterus 0.025 0.00013 0.00026

4.6 URANIUM-233

In the spring of 1961 NFS processed 28 kg of 233U in used pellets. The radiation levels reached 3,000 

mR/hr at 1 inch for 15 g of U3O8 pellets. Twenty operators were exposed to these operations with 

none exceeding the 10 C.F.R. pt. 20 limits; the average exposure for the operation was 377 mrem.

Rubber gloves were worn during the operations in shielded gloveboxes with filtration. Birdcages were 

used for storage of the materials. The exposure rate for the surface of each container did not exceed 

200 mR/hr, and the contamination level did not exceed 500 dpm/100 cm2

. The reading from each 

birdcage did not exceed 1 mR/hr at 1 m (Householder 1963b).

4.7 OCCUPATIONAL EXTERNAL DOSE RECONSTRUCTION ASSUMPTIONS AND 

SUMMARY

Limited film and TLD badge results from the operational era of 1957-1970 are available to develop coworker data in order to determine doses. The minimum detectable limits MDLs of the external 
Document No. ORAUT-TKBS-0043 Revision No. 00 Effective Date: 02/14/2006 Page 42 of 47

dosimetry may be best to use to estimate unmonitored or missed doses. (Table 4-1) The annual 

exposures as determined by the various source terms are estimated and include: 

• Neutron exposure from the (α, n) reaction, (thermal to 2 MeV - penetrating)

• Penetrating extremity exposure from plutonium glove box work, (< 30 keV photons)

• Penetrating and non-penetrating exposure from Uranium metal.

• Penetrating and non-penetrating exposure from contamination. 

• Penetrating and non-penetrating Medical x-ray exposure

Table 4-16 summarizes operational annual external doses to be assigned for monitored and 

unmonitored workers. The annual exposures can be used to estimate doses for unmonitored workers 

or unmonitored periods.

Table 4-16 is to be applied to “Chemical/Production Operators” only for those claims without 

dosimetry information. If dosimetry information is available, the 0.20:1 neutron to photon ratio can be 

applied. For all other radiation workers 10 % of the highest single value in Table 4-16 can be applied.

Table 4-16 Operational external exposure summary. 

Source

Period of 

exposure Exposure category

Exposure 

type Basis

Annual 

exposure 

IREP 

distribution

Unmonitored Workers

Uranium metal 1/1/1957-

12/31/1970

Photons, 30-250 keV, AP 

acute

Penetrating Monte Carlo 

Rectangular ingot

4.16 rem Constant

Uranium metal 1/1/1957-

12/31/1970

Photons, < 30 keV or 

electrons > 15 keV, AP, 

acute

Nonpenetrating10 × penetrating 41.6 rem Constant

Plutonium 

Extremity

1/1/1966-

12/31/1973

Photons, < 30 keV, AP, 

acute (skin on hands and 

forearms)

NonpenetratingDOE 2000 19.125 rem 

ring See 

Table 4-18

Constant

Contaminationa 1/1/1957-

12/31/1970

Photons, 30-250 keV, AP 

acute

Penetrating Table 4-12

(Tables 4-9 to15)

0.064 rem Constant

Contaminationa 1/1/1957-

12/31/1970

Photons, < 30 keV or 

electrons > 15 keV, AP, 

acute

Nonpenetrating10 × penetrating 0.640 rem Constant

Neutron 1/1/1957-

12/31/1970

Neutrons, 0.1-2 MeV, AP 

acute

Penetrating Neutron to photon

& ratio of 0.20:1 

Neutron to 

photon& 

ratio of 

0.20:1b

Constant

Medical x-ray 1/1/1957-

12/31/1970

PA radiographic chest 

exam

Photons, 30-250 keV, AP 

acute & Photons, < 30 keV 

or electrons > 15 keV, AP, 

acute

Penetrating Initial plus one 

examination per 

year

See ORAUTOTIB-0006 

and Table 4-

19

Monitored Workers

Neutron 1/1/1957-

12/31/1970

Neutrons, 0.1-2 MeV, AP 

acute

Penetrating Neutron to 

photon& ratio of 

0.20:1 

Neutron to 

photon& 

ratio of 

0.20:1b

Constant
Document No. ORAUT-TKBS-0043 Revision No. 00 Effective Date: 02/14/2006 Page 43 of 47

Source

Period of 

exposure Exposure category

Exposure 

type Basis

Annual 

exposure 

IREP 

distribution

Medical x-ray 1/1/1957-

12/31/1970

PA radiographic chest 

exam

Photons, 30-250 keV, AP 

acute & Photons, < 30 keV 

or electrons > 15 keV, AP, 

acute

Penetrating Initial plus one 

examination per 

year

See ORAUTOTIB-0006 

and Table 4-

19

Photon Missed 

Dose

1/1/1957-

12/31/1970

Photons, 30-250 keV, AP 

acute

Penetrating Table 4-1 0.02-0.12 

rem 

(MDL/2) 

Dependent

on wear 

period

Lognormal 

GSD 1.52

Photon Missed 

Dose

1/1/1957-

12/31/1970

Photons, < 30 keV or 

electrons > 15 keV, AP, 

acute

NonpenetratingTable 4-1 0.04-0.24 

rem 

(MDL/2)

Dependent

on wear 

period

Lognormal 

GSD 1.52

a. Uranium -233 from 1961-1969.

b. Only to be applied to chemical/production operators.

5.0

 

ESTIMATION OF EXPOSURE TO RESIDUAL ACTIVITY – RESERVED

6.0 ENVIRONMENTAL DOSE - RESERVED 
Document No. ORAUT-TKBS-0043 Revision No. 00 Effective Date: 02/14/2006 Page 44 of 47

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